RAIO-0518-59932, LLC - Submittal of Response to NRC Request for Additional Information No. 386 (Erai No. 9316) on the Design Certification Application

LLC - Submittal of Response to NRC Request for Additional Information No. 386 (Erai No. 9316) on the Design Certification Application

ML18130A811
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Issue date:	05/10/2018
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ML18130A810	List: RAIO-0518-59932, LLC - Submittal of Response to NRC Request for Additional Information No. 386 (Erai No. 9316) on the Design Certification Application
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RAIO-0518-59932 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com May 10, 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 Response to NRC Request for Additional Information No.

386 (eRAI No. 9316) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

386 (eRAI No. 9316)," dated March 13, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The Enclosures to this letter contain NuScale's response to the following RAI Questions from NRC eRAI No. 9316:

03.09.02-50 03.09.02-51 03.09.02-53 03.09.02-56 03.09.02-57 03.09.02-58 The responses to questions 03.09.02-52, 03.09.02-54, and 03.09.02-55 will be provided by September 17, 2018. is the proprietary version of the NuScale Response to NRC RAI No. 386 (eRAI No.

9316). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The proprietary enclosures have been deemed to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.

This letter and the enclosed responses make no new regulatory commitments and no revisions to any existing regulatory commitments.

RAIO-0518-59932 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution:

Omid Tabatabai, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9316, proprietary : NuScale Response to NRC Request for Additional Information eRAI No. 9316, nonproprietary : Affidavit of Zackary W. Rad, AF-0518-59933 Za Zackary W. Rad Director Regulatory Affairs

RAIO-0518-59932 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :

NuScale Response to NRC Request for Additional Information eRAI No. 9316, proprietary

RAIO-0518-59932 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :

NuScale Response to NRC Request for Additional Information eRAI No. 9316, nonproprietary

NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-50 In the response to RAI 8884, Question 03.09.02-2 the detailed vortex shedding evaluations of screened components were discussed. Please add the information to the comprehensive vibration assessment (CVAP) report TR-0716-50439. Without the information, the staff cannot reach a reasonable assurance finding on the structural integrity of the reactor internals components to withstand the adverse effects of vibration.

NuScale Response:

In the response to RAI 8884, Question 03.09.02-2, discussion of the acoustic resonance evaluations of screened components was provided.

This RAI question requested that this information be added to the CVAP technical report TR-0716-50439.

NuScale has updated TR-0716-50439 Sections 2.3.3 through 2.3.6 to include discussion of the evaluations for acoustic resonance of screened components, as provided in the response to RAI 8884, Question 03.09.02-2.

Impact on DCA:

Sections 2.3.3, 2.3.4, 2.3.5 and 2.3.6 of TR-0716-50439 have been revised as described in the response above and as shown in the markup provided with this response.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 24 cylinder. Fluid enters the upper riser section at the transition and turns 180 degrees over the upper edge of the riser.

The upper riser section is susceptible to parallel flow TB due to the upward flow inside the riser and the downward flow on the riser exterior. The upper riser section is not susceptible to FEI, AR, gallop, or flutter as it is an open cylinder. The upper riser section directs the fluid flow and does not cross the flow path, precluding it from a VS susceptibility. Flow in the region above the riser is fully turbulent. This region does not represent a cavity, where AR could develop, because it contains the CRDS, ICIGT, hot leg thermowells, hanger braces, as well as the SG tubes. Therefore, FIV due to mechanisms other than TB are not credible for the upper riser section.

Figure 2-13 Upper riser section with characteristic flow direction noted

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 28 2.3.3.6 Upper Riser Hanger Brace As shown in Figure 2-16, the upper riser hanger assembly connects the upper riser section to the pressurizer baffle plate. Fasteners are used to attach the hanger ring to the baffle plate, such that there is no flow past or above this part. However, similar to the ICIGT and CRD shaft, the upper riser hanger braces experience cross flow as the hot leg fluid turns from the upper riser into the SG region. As the flow area above the upper riser and below the integrated steam plenum is not a cavity, generation of an AR condition due to the vortices shed from the hanger assembly is not possible. Based on this, the upper riser hanger braces are susceptible to TB and VS. No other FIV mechanisms are credible for this region.

Figure 2-16 Upper riser hanger assembly 2.3.4 Lower Riser Assembly The lower riser assembly includes the lower riser section, the upper core plate, CRA guide tubes, CRA guide tube support plate, and CRD shaft supports. The portions of the lower riser assembly that screen for FIV are identified in the following subsections.

2.3.4.1 Lower Riser Section The lower riser section is the cylindrical section in the lower riser assembly, as shown in Figure 2-17. This section transfers the loads from the slip joint and the guide tube

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 30 with a flexible boundary. There is no cavity region in the CRAGT assembly where AR could form. The CRAGT assembly is designed to allow flow to pass in and out of the guide tube. There are no flow-occluded regions and any vortices that form are dissipated by the turbulent flow. Using the screening criteria, the CRAGT is not susceptible to the FIV phenomena, other than VS and TB.

Figure 2-18 Control rod assembly guide tube assembly 2.3.4.3 Control Rod Assembly Guide Tube Support Plate The CRAGT support plates are located above the CRAGT assembly, as depicted in Figure 2-17. Similar to the CRAGT assembly, the support plate is subject to turbulent flow and also represents a bluff body subject to cross flow. Further, there is no cavity region downstream of the CRAGT support plate where AR could form. As shown in Figure 2-15, the downstream region contains the CRDS, CRDS supports and ICIGT.

Therefore, TB and VS are applicable mechanisms. Other mechanisms are not considered credible.

2.3.4.4 Upper Core Plate The upper core plate functions in conjunction with the core support assembly to align and support the reactor core system. The upper core plate is welded to the bottom of the lower riser, as depicted in Figure 2-17. Four lock plate assemblies align the lower riser assembly with the core support structure and secure the two assemblies together. The upper core plate is subject to turbulent flow and also represents a bluff body subject to

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 32 Figure 2-19 Core support assembly 2.3.5.1 Core Barrel The core barrel is a large cylinder designed to carry the core support loads and separate the down-flowing fluid from the fuel, as denoted in Figure 2-19. The core barrel is susceptible to TB using the screening criteria. Using the screening criteria, the core barrel is not susceptible to the other FIV phenomena.

2.3.5.2 Upper Support Block The upper support block is attached to the core barrel and opposes the fluid as it travels through the downcomer, as shown in Figure 2-19. There is no cavity region downstream of the upper support block where AR could form. This block is susceptible to VS and TB phenomena. Using the screening criteria, the upper support block is not susceptible to the other FIV phenomena.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 33 2.3.5.3 Core Support Block The core support blocks are located in the downcomer. This feature transfers the dead weight and accident loads from the lower core plate to the lower head of the RPV, as denoted in Figure 2-24. There is no cavity region downstream of the core support blocks where AR could form. This block opposes the fluid flowing through the downcomer into the lower plenum and is susceptible to VS and TB phenomena. Using the screening criteria, the core support block is not susceptible to the other FIV mechanisms.

2.3.5.4 Belleville Spring The Belleville springs, shown in Figure 2-20, transfer the dead weight, seismic and accident loads from the lower core plate to the lower head of the RPV. Each core support block location contains two Belleville springs to adequately support the core during seismic events. The smaller Belleville spring is located between the retaining nut and the lower core plate. The larger Belleville spring is located under the lower core plate within the core support block. The Belleville springs do not experience any cross flow since they are contained above and within the core support block (i.e. no downstream flow path). Although there will be some small clearance flow paths between the washers that make up the Belleville springs, there is not expected to be appreciable flow through these gaps because the center of the Belleville springs is an annular cavity with minimal to no outflow and thus there is minimal to no pressure drop possible across the Belleville washers. Further, the inside of the springs contain either a retaining pin (lower spring) or the core support block pin (upper spring) which will impede flow into the cavity. Based on these considerations, leakage flow is not judged to be possible in this region. The Belleville springs do experience parallel flow and are therefore susceptible to turbulent buffeting. Belleville springs are not susceptible to the other FIV phenomena.

Figure 2-20 Core support block and Belleville spring

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 34 2.3.5.5 Reflector Block The reflector blocks are aligned by pins and stacked on the lower core plate inside the core barrel, as denoted in Figure 2-19. Primary coolant flows through small channels in the reflector and the inner surface of the reflector is subject to turbulent core flow. The only mechanism applicable to this component is TB. Other mechanisms are not credible due to the flow conditions and component geometry.

2.3.5.6 Lower Core Plate The lower core plate is located below the fuel assemblies, as depicted in Figure 2-19.

The lower core plate is subject to turbulent flow and also represents a bluff body subject to cross flow. There is no cavity region downstream of the lower core plate where AR could form. The structures downstream of the lower core plate are primarily narrow flow channels composed of fuel assemblies. Therefore, TB and VS are applicable mechanisms. Other mechanisms are not considered credible including acoustic resonance, which cannot form in the highly turbulent conditions downstream of the lower core plate. Similar to the upper core plate, the lower core plate is significantly thicker and stiffer than the CRAGT support plate and experience similar flow velocities; therefore, the response of the lower core plate to these mechanisms is expected to be bounded.

2.3.5.7 Fuel Pin Interface The fuel top and bottom nozzles interface with fuel pins that are installed in the upper and lower core plates, as shown in Figure 2-21. This interface is a location for impact primarily driven by the TB of the fuel assembly. The fuel pin is not directly subject to other FIV mechanisms based on component geometry and flow conditions.

Figure 2-21 Typical fuel nozzle-to-fuel pin interface 2.3.6 Other Reactor Vessel Internals

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 35 There are other design features located in the primary and secondary coolant flow paths that require FIV screening. These consist of instrumentation and system connections that support interfacing systems. For the NPM design, the majority of the RPV connections are located in the steam space of the pressurizer. This region is always steam and is not exposed to high average flow rates, even under transient and accident conditions, due to the very large cross sectional area on the region. Therefore, for the majority of the instrument and interfacing system connections in the NPM design, degradation due to FIV is not credible. The instrumentation and system connections that screen for FIV are identified in the following subsections.

2.3.6.1 Pressurizer Spray Reactor Vessel Internals The pressurizer spray lines and nozzles are attached to a nozzle in the upper RPV head and extend downward into the steam region of the pressurizer (Figure 2-22). Fluid is pumped through these components to provide the pressurizer spray. Jet flow fluctuations do not occur at the pressurizer spray nozzle because the chemical and volume control system (CVCS) is a forced flow system. The flow is subsonic and does not undergo phase change upon exiting the nozzle. Therefore, flow vortices are not generated. These lines are susceptible to TB as there is a large pressure loss across the nozzle and flow rates through the spray line piping are turbulent when spray flow is used. Using the screening criteria, the pressurizer spray lines and nozzles are not susceptible to the other FIV mechanisms.

Figure 2-22 Pressurizer spray reactor vessel internals 2.3.6.2 Chemical and Volume Control System Injection Reactor Vessel Internals The CVCS inlet line is routed from a flange in the RPV wall into the downcomer (Figure 2-23). The line continues through the upper riser section into the up-flowing region above the core. The line is susceptible to VS and TB in both regions. Using the screening criteria, the CVCS injection RVI is not susceptible to other FIV phenomena.

NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-51 In the response to RAI 8884, Question 03.09.02-3, the staff finds that removal of the previously planned control rod assembly guide tube (CRAGT) FIV testing needs quantitative detailed justification. The very low cited margin of the CRAGT to flow-induced vibration (FIV) induced wear is based on an unverified assumed forcing function (NuScale referred to their assumption discussion in EC-A023-3535, Rev. 0, RVI Turbulent Buffeting Evaluation). Explain quantitatively how fluid flow estimates within and around the CRAGT (particularly through the CRAGT side holes), along with flow-induced forcing functions and vibration response are confirmed to be conservative.

Add the information to the comprehensive vibration assessment (CVAP) report TR-0716-50439.

NuScale Response:

The TR-0716-50439, Section 3.0 discusses the conditions for validating FIV predictions with tests. Section 3.1 establishes the criterion for testing to validate turbulent buffeting. For turbulent buffeting, when a large margin of safety exists for detrimental effects from fatigue over the design life of the component, no testing is necessary. The fatigue evaluation for the CRAGT shows negligible fatigue usage over the design life due to turbulent buffeting induced vibrations.

The integrity of the component is not challenged by vibrations, regardless of reasonable inaccuracy in the assumed forcing function for the CRAGT. Because the margin to a cumulative usage of one for impact and fatigue exceeds nine orders of magnitude, the CRAGT was not selected for FIV testing.

Although wear was estimated in the RVI Turbulent Buffeting Evaluation, EC-A023-3535, quantitative wear estimates are not part of the CVAP. Regulatory Guide 1.20 and SRP Section 3.9.2 list guidelines that, when met, constitute satisfaction of GDCs 1 and 4 with respect to flow induced vibration effects. These guidelines do not include the evaluation of wear or acceptance criteria for wear. The wear evaluation was conducted to address the business risk of having to replace CRAGTs during the life of the plant and not to address a regulatory requirement.

NuScale Nonproprietary The CVAP, TR-0716-50439, does not identify wear as a parameter for the Vibration Measurement Program. The inspection for wear of components is a part of the Vibration Inspection Program, but is not associated with pretest predictions for the amount of wear. The inspection for vibration induced damage provides a secondary confirmation of the FIV integrity of the NPM components.

Wear, if present, is detected through inservice inspection of the reactor vessel internals in accordance with Section XI of the ASME BPVC. Inspections of the CRAGT interfaces are described in FSAR Tier 2, Chapter 5.2, Table 5.2-7. These inspections ensure that the plant is not operated with components that have experienced excessive wear.

The use of the annular/channel flow forcing function has been quantitatively justified as bounding in the RVI Turbulent Buffeting Evaluation, and the cited unverified assumption justified as not requiring further verification. The selected PSD is higher across the frequency range than the potential alternative PSDs, including those for cross flow. A comparison of the annular/

channel and cross flow PSDs is performed at an RCS flowrate that is 3 percent higher than the maximum design flow at full power from FSAR Table 5.1-2. The PSD for annular/channel flow is larger across the frequency range than the cross flow PSD. This provides confidence that the evaluation for wear is bounding based on the PSD input.

There are additional conservative methods in the wear estimate. The evaluation estimates the maximum RMS vibration amplitude for the CRAGT. The maximum RMS vibration amplitude occurs in the middle of the CRAGT. This amplitude is then used to calculate the normal force for a tube whirling in an oversized hole, which represents the interface between the CRAGT and the CRAGT support. This is a conservatism in the analysis method because the RMS amplitude at the supported terminal ends of the CRAGT is lower than the maximum amplitude in the middle of the span.

Additionally, the CRAGT is warmer than the exterior of the lower riser structure during normal operation. The CRAGT support and the upper core plate are both attached to the interior of the lower riser. The distance between these CRAGT terminal supports depends on the temperature of the riser wall due to thermal expansion. Since the CRAGT is hotter than the riser wall during normal operation, it will grow axially more than the distance between the supports will increase, causing a reaction force to develop at the interface between the CRAGT and the CRAGT support. Thermal expansion of the CRAGT into its supports has not been considered in this estimate, but is expected to reduce relative motion at the interface.

In summary, the CVAP analysis with regard to the CRAGT demonstrates large margins of safety for fatigue over the design life of the component. No further validation of this assessment is required because the margin of safety exceeds nine orders of magnitude. Significant wear of the CRAGT is not expected in service. If wear occurs, it is detected by inservice inspection programs that prevent operation with degraded components. Although not part of the CVAP scope, the wear evaluation of the CRAGT is accomplished using conservative methods that overestimate the potential for wear of this component.

NuScale Nonproprietary Because wear evaluations are not within the scope of the CVAP, the TR has not been revised based on this response.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary Response to Request for Additional Information Docket No.52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-53 In the response to RAI 8884, Question 03.09.02-5, the table of velocities added to the markup CVAP report is incomplete. The table is limited, and does not include velocities for all evaluated components and FIV mechanisms. As originally requested, add detailed velocity information to the CVAP report for all evaluated components and FIV mechanisms, including fluid-elastic instability (FEI), VS, turbulence buffeting (TB), and acoustic resonance (AR). Also include velocities computed from thermal hydraulic analysis, being sure to include those used for steam generator (SG) tube and SG support bar VS and FEI, as well as DHRS AR assessments.

Include any available bias errors and uncertainties based on existing thermal hydraulic test data. Without the information, the staff cannot reach a reasonable assurance finding on the structural integrity of the reactor internals components to withstand the adverse effects of vibration because the velocity is a key parameter in the flow-induced vibration analysis and the application must contain a level of design information sufficient to enable the staff to judge the applicants proposed design.

NuScale Response:

A detailed table of velocities has been added to Section 3.1.2.3 of the CVAP technical report (TR-0716-50439). The table provides flow velocities used for flow-induced vibration (FIV) analyses of the evaluated components and FIV mechanisms. Derived velocities unique to a certain FIV phenomenon, such as the convective velocity in turbulent buffeting calculations, are discussed qualitatively in Section 3.2 of the CVAP technical report.

Additionally, Section 3.1.2.1 of the CVAP technical report has been supplemented to include information on the biases used in the thermal-hydraulic calculation of the bounding maximum design flow rate. The relationship between bias and uncertainties in the thermal-hydraulic data is discussed.

NuScale Nonproprietary Impact on DCA:

Section 3.1.2 and Table 3-5 of TR-0716-50439 have been revised as described in the response above and as shown in the markup provided with this response.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12

© Copyright 2018 by NuScale Power, LLC 45 verification is not required for each mechanism due to differences in the safety margins.

For example, the ICIGT frequencies require verification to validate the VS analysis inputs, but validation of this input for the TB analysis is not required. However, if the test data is available for a component, it is used for post-test evaluation of all mechanisms for that component, to the extent practical.

Table 3-3 Structural natural frequencies and mode shapes input summary Component Analysis Category Safety Margin Helical SG tubing FEI Note(2)

((2(b),(c),ECI TB ((

}}2(b),(c),ECI VS Note(1), (3)

((

}}2(b),(c),ECI ICIGT VS Note(1)

((

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

((

}}2(b),(c),ECI CRD shaft VS

((

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

((

}}2(b),(c),ECI Note(s) for Table 3-3:

1.

Mode shape and damping ratio are used in the SG tube and ICIGT VS evaluations. These inputs are not required for VS evaluation of other components.

2.

FEI safety margin is reported for the limiting SG tube column (column 21), assuming active support and heavy tube conditions.

3.

VS safety margin is reported for the limiting SG tube column (column 1), assuming active support and light tube conditions. 3.1.2 Flow Velocity 3.1.2.1 Steady State Velocity Analysis Appropriate representation of the flow conditions is essential to performing the analysis of all FIV phenomena. The flow velocities in the axial/parallel and crossflow directions relative to the normal axes of each component are needed as analytical inputs. Due to the importance of generating appropriate flow rates, RCS and secondary-flow rates for the various regions of the NPM are determined and verified using three different analytical methods: hand calculation, thermal-hydraulic (TH) analysis, and computational fluid dynamics (CFD) analysis. Additionally, separate effects tests have been performed to validate the significant inputs in the TH modeling, such as core and SG form losses and SG performance, which is a significant contributor to the natural circulation buoyant force. The TH analysis provides validated maximum design flow rate results based on testing. The CFD average flow velocities (Reference 8.1.12) and hand calculations provide additional assurance of the validated flow rates. Maximum design flow rates (or higher) are used in all FIV evaluations. Maximum design flow represents the highest expected primary coolant flow rate used in design and safety analyses and is the licensing basis flow rate; operation above this flow rate is not permitted without evaluation against safety analysis limits.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 46 Several biases are applied in the TH analysis to ensure the calculation of a bounding maximum design flow rate. The steam generator and fuel assembly loss coefficients are biased based on the uncertainty in the loss coefficient correlations. Other pressure losses are decreased to be bounding. Factors such as core bypass flow, and steam generator heat transfer are biased to provide a bounding primary coolant flow rate. Steam generator loss coefficient bias: The best-estimate loss coefficient for the steam generator is decreased by ((

}}2(b),(c),ECI for the maximum design flow calculation. This accounts both for error in the loss coefficient correlation, as determined by comparison to test data, and test data measurement uncertainty.

Fuel assembly loss coefficient bias: ((

}}2(b),(c),ECI is subtracted from the overall fuel assembly loss coefficient to bound loss coefficient correlation confidence and experimental uncertainty. The bias value is based on pressure drop testing of a full-size prototype of the NuScale fuel assembly.

Reactor coolant system form loss coefficient bias: For the regions in the RCS loop except for the core and steam generator, the best-estimate form loss coefficients are reduced by ((

}}2(b),(c),ECI for calculation of the maximum design flow. Form loss coefficients for most of the components in the RCS loop are based on empirical correlations. Using a comparison of test data to an empirical correlation, a

((

}}2(b),(c),ECI reduction is appropriate for regions outside the SG and core. The sum of the SG and core form losses is greater than (( 
}}2(b),(c),ECI of the total pressure loss in the RCS loop, making the choice of form loss coefficient elsewhere of less significance.

Core bypass flow bias: To address uncertainty in the fraction of the total RCS flow rate that bypasses the core, ((

}}2(b),(c),ECI of the total RCS flow is added to the best estimate core bypass flow for the maximum design flow conditions.

Steam generator heat transfer bias: Thermal conductivity of the steam generator tube wall material is decreased by a factor of ((

}}2(b),(c),ECI for the maximum design flow calculation. Changing the tube wall conductivity in this manner has a significant effect on the value of the overall heat transfer coefficient. The equivalent secondary side film coefficient variation produced by the change in wall conductivity bounds the average inaccuracies in the TH analysis heat transfer predictions as compared to the steam generator test data.

For most FIV mechanisms, the free-stream velocity is the required input. Free-stream velocity is the incident velocity before it impacts a susceptible target. However, in the FIV analyses it is typically assumed to be the flow velocity as it passes the target or constriction. The flow area upstream of the constriction is larger than the flow area associated with the constriction; therefore, the average free-stream velocity is lower than the average velocity at the target. While there may be local variations in the free-stream velocity, it is not credible that these variations would occur uniformly at the target face at a velocity higher than the average velocity through the constricted area.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 47 For components where it is bounding to maximize the convective velocity, the TB analyses are performed using the maximum CFD flow rates calculated at a region. While these flow rates are not specifically validated, they significantly bound the average constriction flow velocities, which bound the average free-stream flow velocities; and therefore do not require analytical or experimental validation. 3.1.2.2 Transient Velocity Analysis Transient analysis is performed to determine pressures, temperatures, flow velocities, and static quality for the various NPM regions to support ASME stress analysis. The NPM typically operates at steady-state, full-power conditions. At lower reactor power levels, natural circulation flow decreases due to the lower heat addition from the core. Based on a review of time-history analyses, there are no normal operating transient responses that result in higher flow rates than are achieved at maximum design full-power conditions. Normal operating transients, such as load following, ramp decreases in reactor power, and reactor trips, result in decreases in flow velocities relative to full-power operating conditions due to the lower heat addition from the core. For conservatism, flow velocities 5% or greater than the maximum design flow value should be used for all FIV mechanisms except for turbulent buffeting. For turbulent buffeting, maximum design flow velocities are acceptable. 3.1.2.3 Velocities Used in Flow-Induced Vibration Analysis To determine the crossflow velocities for components subject to a combination of crossflow and axial flow, since bounding estimates of cross flow velocities can be assumed while still showing acceptable FIV performance, it can be conservatively assumed that the crossflow on any component is equal to the calculated total flow velocity for that component. This method overestimates the forcing function of the vibration. Based on the low reactor coolant flow rate for the NuScale design, this approach is a conservative estimate that produces acceptable margin. Table 3-4 identifies the methods of obtaining the velocities that are used in the FIV analyses and which flow rate analysis derives them. Bounding flow conditions are used in all analyses. Testing to validate the velocities is not necessary because these FIV inputs have already been validated with separate effects testing, and one or more analytical method and bounding maximum design flow velocities are used.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 48 Table 3-4 Flow conditions input summary Analysis Category Assumed Conditions Analysis Method FEI Maximum design flow - average velocity CFD Note 1 VS Maximum design flow - average velocity CFD Note 1 Design flow at 102% - average velocity assuming low SG superheat performance TH Note 4 AR Maximum design flow - average velocity CFD Design flow at 102% - average velocity assuming low SG superheat performance TH Note 4 F/G Maximum design flow - average velocity CFD LFI None Note 2 None TB Maximum design flow - average or maximum velocity Note 3 CFD Note(s) for Table 3-4:

1.

For the VS and FEI analysis of the SG tubes, CFD flow rate is modified to represent velocity in the minimum flow area.

2.

LFI confirmation is by prototype testing only.

3.

Either average or maximum velocities are chosen for each component in TB analysis based on achieving the most bounding convective velocity for the purpose of impact and fatigue evaluations.

4.

For the evaluation of AR and VS mechanisms for components exposed to secondary coolant flows, the TH flow is used. CFD analysis is not performed to characterize secondary side flow. Table 3-5 lists velocities used in the analyses of components subject to FIVrepresentative average velocities based on CFD analysis at the maximum design flow. The analysis methods that produce these velocitiescategories and components that use the CFD results are identified in Table 3-4, except as noted otherwise below. Table 3 5 Maximum design flow velocities based on CFD Flow Region Average Velocity (in/s) Around/Through CRAGTs ((

}}2(b),(c),ECI CRAGT Support

((

}}2(b),(c),ECI Bottom of Conic Riser Transition

((

}}2(b),(c),ECI CRD Shaft Support

((

}}2(b),(c),ECI Upper Riser

((

}}2(b),(c),ECI Flow Over the Top of the Upper Riser

((

}}2(b),(c),ECI Top of Conic Downcomer Transition

((

}}2(b),(c),ECI Bottom of Conic Downcomer Transition

((

}}2(b),(c),ECI Downcomer, around Core Barrel

((

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

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 49 Table 3-5 Velocities used in FIV analyses Analysis Category Component Velocity (in/s) FEI Note 1 Helical SG tubing (( VS Note 1 Helical SG tubing SG tube support cantilever RCS hot region thermowell RCS cold region thermowell CNTS steam thermowell CNTS feedwater thermowell Control rod drive shafts CRD shaft support Control rod assembly guide tubes CRAGT support Upper riser hanger brace CVCS Injection RVI (in downcomer) In-core instrument guide tubes AR Note 4 DHRS steam line tee DHRS condensate line tee Reactor recirculation valve nozzle Flowmeter port F/G Note 1 SG tube support cantilever

}}2(b),(c),ECI LFI None, LFI confirmation is by prototype testing only TB Helical SG tubing, primary flow

(( Helical SG tubing, secondary flow (steam) Helical SG tubing, secondary flow (liquid) SG inlet flow restrictor Core barrel CRAGT inner diameter CRAGT outer diameter CRAGT support CRD shaft CRD shaft support Flow diverter Lower ICIGT

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

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 50 Analysis Category Component Velocity (in/s) Upper ICIGT (( Injection line, downcomer Injection line, downcomer, interior Lower core plate Lower riser inner diameter Lower riser outer diameter Upper core plate Upper riser inner diameter Upper riser outer diameter

}}2(b),(c),ECI Notes for Table 3-5:

1.

((

}}2(b),(c),ECI margin is included in these values for transient velocity changes

2.

Primary side gap velocity based on a minimum cross-sectional flow area

3.

Component of this velocity perpendicular to the tubes is used in the analysis

4.

((

}}2(b),(c),ECI is added to these values in the analysis to account for transient velocity changes

5.

Velocity is from TH analysis

6.

Velocity is hand-calculated

NuScale Nonproprietary Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-56 In the response to RAI 8884, Question 03.09.02-13, NuScale provided quantitative criteria for evaluating structures for fatigue due to turbulence buffeting and markup changes to the CVAP report. The staff requests that the applicant add details on the turbulent buffeting evaluations of components with lowest resonance frequencies less than 200 Hz to the CVAP report. Without the information, the staff cannot reach a reasonable assurance finding on the structural integrity of the reactor internals components to withstand the adverse effects of vibration. NuScale Response: Section 3.2.3 of the CVAP technical report (TR-0716-50439) addresses turbulent buffeting (TB) analysis. Further discussion on how components are selected for TB analysis has been added to this section of the technical report. The TB evaluation methodology has also been supplemented to clarify calculation of the component mean square responses and impact fatigue effects for components with fundamental frequencies below 200 Hz. Impact on DCA: Section 3.2.3 of TR-0716-50439 has been revised as described in the response above and as shown in the markup provided with this response.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 55 3.2.3 Turbulent Buffeting Although the random vibration due to TB is of much smaller amplitude than that experienced in FEI or VS, it is important because it exists whenever there is turbulent flow over a susceptible component. Turbulent buffeting vibration can be induced in parallel flow, axial flow, or crossflow. Turbulence-induced vibration occurs due to a fluctuating pressure in the flow. The fluctuating pressure is quantified by a forcing function, which is characterized by its power spectral density (PSD) and correlation function. Three significant parameters are needed to characterize the PSD and correlation function: convective velocity, correlation length, and the PSD. The convective velocity as implemented in the coherence function, describes the phase relationship of the forcing function at two different points on the surface of the structure. From a physical standpoint, the convective velocity is the velocity at which eddies are swept downstream, thus causing a phase shift in the fluctuating pressure between two points. Based on literature review, convective velocity can be a maximum of the free-steam velocity and a minimum of 0.5 times the free-stream velocity for parallel or axial flow. The limiting convective velocity within the range is selected when evaluating the coherence function. The correlation length is equivalent to the scale of the largest turbulent eddy, which is a measure of the longest distance over which the velocity at two points of the flow field is correlated. For each component, the flow is characterized as parallel, cross, or axial, and the characteristic length is determined. A bounding correlation length range is determined based on literature review. Similar to the convective velocity, the range of correlation lengths is evaluated to ensure a bounding value is selected for each component forcing function. The PSD is determined to support characterization of the forcing function. The PSD as applied to flow-induced turbulent response refers to the energy distribution of a variable as a function of frequency. Many components can be approximated as cylinders in cross flow or annular flow. For these idealized arrangements, bounding PSDs based on literature are used. Table 3-9 summarizes the PSD types that are applied to components susceptible to TB. The overall analytical uncertainty for the SG tubes is judged to be the highest relative to other susceptible components; therefore, prototype testing is performed to verify the adequacy of the PSDs used for the SG tubes. In the TB evaluations, the vibration amplitude is not calculated for components with a first mode frequency greater than 200 Hz because vibration amplitudes are insignificant with high structural frequencies. Components with lower first mode frequencies experience higher root mean square (RMS) vibration amplitudes, which translate into more limiting fatigue results. Vibration amplitude is also not determined for components that based on flexibility and flow conditions can be shown to be bounded by the analyzed components. By design, supports are thicker and less flexible than the components they support; hence, the component is analyzed for turbulent buffeting vibrations. The results generally bound the performance of the support. Therefore, no

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 56 PSD is specified in Table 3-9 for certain components identified as susceptible to TB in Table 3-1. Table 3-9 Turbulent buffeting power spectral density inputs used in analysis Flow Type and Component Shape Applicable Components Literature PSD Used Parallel flow over isolated cylinder (( Equation 3-2 Tube bundle cross-flow Equation 3-3 Bounding annular flow Equation 3-1

}}2(a),(c),ECI Note(s) for Table 3-9:

1.

Testing is performed to verify that the pressure PSD due to internal two-phase flow is sufficiently bounded by the pressure PSD for single-phase flow to justify use of the single-phase pressure PSD used in the analysis. Equation 3-1 is from Reference 8.1.3. It provides a PSD for components with annular and axial flow velocities. It is applied to some components that experience crossflow, and this simplification is bounding based on the frequencies and characteristic lengths of the analyzed components. Due to the relatively low flow velocities, the reduced frequency for some components is larger than five. For those cases, the PSD is evaluated with a reduced frequency of five to provide bounding results.

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 58 ( ) 4 3.5 5.3 10 0.4 p G f x F F

= where: ( ) P f G = PSD of the turbulent pressure as a function of modal frequency (psi2/Hz) and F = Reduced frequency (-). The structural response due to the turbulence is calculated using the inputs that have been discussed. Equations to determine the root mean square (RMS) response are assigned based on the direction of the flow and the dimension of the structure, using the appropriate PSDs, damping ratios, modal analysis results, and flow characteristics for the analyzed components. The mean square response of the SG tubes is derived using Equations 9.4 and 9.40 of Reference 8.1.3. Equations 8.45, 8.46, and 8.47 of Reference 8.1.3 are the basis for the RVI mean square response calculation. Using the RMS response, degradation mechanisms associated with impact and vibration fatigue are evaluated. Components with less than one inch of separation are evaluated for impact. Components with fundamental frequencies less than 200 Hz and those whose response cannot be bounded by nearby components exposed to similar turbulent conditions are evaluated for fatigue. Table 3-9 lists the components analyzed for turbulent buffeting. Each has a fundamental frequency below 200 Hz. After the RMS response is calculated as discussed above, components undergo further separation screening: Tthe acceptance criterion for a component remaining separated from an adjacent component is that the clearance is greater than 5 RMS deflection. For components that do not meet the criterion, the surface stress is calculated and an impact fatigue assessment is performed. Surface stress is calculated using a semi-empirical approximation from Reference 8.1.3 for a vibrating tube in a loose hole (Equation 3-4). The impact frequency for the steam generator tubes is the average of the first ten modal frequencies, because these modes contribute the majority of the RMS response. For RVI evaluations, the crossing frequency is used. 1 5 4 2 2 max 3 e i rms E M f y S c D

=

Equation 3-4 where: rms S = Surface stress due to impact (psi) c = Contact stress parameter, page 357 of Reference 8.1.3 (-) E = Elastic modulus (psi)

NuScale Comprehensive Vibration Assessment Program Technical Report TR-0716-50439-NP Draft Rev. 12 © Copyright 2018 by NuScale Power, LLC 59 e M = Effective mass, for a tube usually taken as 2/3 of the mass of the two adjacent spans (lbf-s2/in) if = Impact frequency (Hz) max y = Maximum RMS vibration amplitude (in) D = Tube outer diameter (in) An alternating impact stress is calculated as one-half of the surface stress, and used with a fatigue curve based on RMS stress (similar to Figure 11.6 of Reference 8.1.3) to obtain a usage factor. In addition to the impact fatigue, TB can cause fatigue due to vibration stresses. However, due to the very low vibration amplitudes and alternating stresses, the vibration stresses do not result in fatigue usage for any component susceptible to TB. Out of the components analyzed for TB, four components are shown to impact their adjacent supports and are discussed below. Impact between the SG tubes and SG tube support is predicted to occur, and the fatigue usage due to vibration and impact is calculated to be ((

}}2(b),(c),ECI which provides a margin of safety of (( 
}}2(b),(c),ECI This result demonstrates adequate SG tube performance for the component design life, subject to verification testing of analytical inputs and safety margin as identified in Table 3-10.

Impact between the ICIGT and the CRD shaft supports, between the CRD shaft and CRD shaft supports, and between the CRAGT and the guide tube support is predicted to occur. For all three impact pairs, the results show that the RMS vibrations do not result in fatigue usage due to the displacements themselves or due to impact. These results are due to the very low alternating stresses generated from the TB, which can be primarily attributed to the low-flow velocities. Table 3-10 provides an overview of the analysis results and required testing.

NuScale Nonproprietary Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-57 In the response to RAI 8740, Question 14.02-1, NuScale revised the DCD Tier 2, Chapter 14.2 descriptions for Tests #44, #45, and #72 from startup testing to factory tests or separate effects tests. DCD Tier 2, Section 14.2 is for preoperational and initial startup testing. As indicated in DCD Tier 2, Section 14.2, NuScale is not planning to conduct preoperational testing for reactor internals. Startup testing is defined as testing performed in the assembled plant after fuel load. However, NuScale listed the following factory tests or special effects tests in Chapter 14.2: Test #44: Control Rod Drive System Flow-Induced Vibration Test performed during factory testing Test #45: Reactor Vessel Internals Flow-Induced Vibration Test (only in-core instrument guide tubes) performed at the factory Test #72: Steam Generator Flow-Induced Vibration Test [These are actually the separate effects tests on steam generator tube columns (but with only a limited number of tube columns and possibly no flow inside the SG tubes), and steam generator inlet flow restrictors] These are tests that will be performed in a laboratory or factory, and not in the assembled plant as a hot functional test or initial startup test. Therefore, relocate these tests from DCD Tier 2, Section 14.2, or provide justification for including these tests in Chapter 14.2. NuScale Response: The NRC RAI 8740 Question 14.02-1 noted that previously approved design certifications included design features, which are new and unique and have not been tested previously, in their initial test program (FSAR Chapter 14, Section 14.2) first-of-a-kind tests that were only to be performed a limited number of times. This RAI question requested NuScale to similarly identify such tests. The FSAR Section 14.2 changes made in the response to RAI 8740, Question 14.02-1 included additional content in Section 14.2.3.3, Testing of First-of-a-Kind Design Features, as well as revised descriptions for Tests #44, 45, and 72. This additional content identifies that the CVAP

NuScale Nonproprietary program requires Tests #44 and #45 which are factory tests, and Test #72 which is a separate effects test performed at a test facility. These tests are performed per Table 4-1 of the NuScale Comprehensive Vibration Assessment Program Technical Report, TR-0716-50439. Tests #44,

1. 45 and Test #72 require visual examination of the CRDS, RVI, and SG components, respectively, as specified in Table 5-1 of TR-0716-50439. These tests are coordinated with startup Test #108.

Tests #44, #45, and #72 are discussed in Section 14.2 in order to identify how flow-induced vibration testing (normally performed during the preoperational testing described in Section 14.2) is performed for the NuScale design. Each of these three test abstracts states that flow-induced vibration testing is not performed during the NuScale preoperational test phase. Impact on DCA: There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary Response to Request for Additional Information Docket No. 52-048 eRAI No.: 9316 Date of RAI Issue: 03/13/2018 NRC Question No.: 03.09.02-58 10 CFR 50, Appendix A, GDC 4 requires structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. DCD, Tier 2, Table 1.9-2, comments for RG 1.20 state the following: The aspects of this RG that mandate start-up tests or inspections for non-prototype reactors are applicable as these tests must be accommodated in the start-up of each installed NuScale Power Module. The remainder is applicable to the COL applicant, and only specific COL applicants which are for prototype reactors. The comments are not clear. Provide clarification on these statements and the exceptions to RG 1.20. Revise the DCD if necessary. NuScale Response: This question requested clarification of the conformance status and comments for Regulatory Guide 1.20, Comprehensive Vibration Asessment Program for Reactor Internals During Pre-operational and Initial Startup Testing, in Table 1.9-2 of the FSAR. NuScale has amended the entry for Regulatory Guide 1.20 in Table 1.9-2 of the FSAR to indicate that NuScale conforms with RG 1.20. The comments explain that 'The first operational NPM is classified as a prototype in accordance with RG 1.20. Thus, the portions of this RG that apply to prototype reactors are applicable to the first operational NPM. After the first NPM is qualified as a valid prototype, subsequent NPMs are classified as non-prototype category I and the non-prototype portions of the RG apply. The identification of departures from RG 1.20 is the responsibility of the COL applicant.'

NuScale Nonproprietary Impact on DCA: The FSAR Tier 2, Section 1.9 Table 1.9-2 has been revised as described in the response above and as shown in the markup provided with this response.

NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Tier 2 1.9-5 Draft Revision 2 RAI 01-1, RAI 02.03.01-5, RAI 03.09.02-58, RAI 05.02.03-13, RAI 05.03.01-3, RAI 05.04.02.01-11, RAI 06.01.01-8, RAI 06.01.01-9, RAI 08.01-1, RAI 08.01-1S1, RAI 08.02-4, RAI 08.02-6, RAI 08.03.02-1, RAI 09.02.06-1, RAI 14.02-7 Table 1.9-2: Conformance with Regulatory Guides RG Division Title Rev. Conformance Sta-tus Comments Section 1.3 Assumptions Used for Evalu-ating the Potential Radiologi-cal Consequences of a Loss of Coolant Accident for Boiling Water Reactors 2 Not Applicable This guidance is only applicable to BWRs. Not Applicable 1.4 Assumptions Used for Evalu-ating the Potential Radiologi-cal Consequences of a Loss of Coolant Accident for Pressur-ized Water Reactors 2 Not Applicable This RG pertains to existing reactors; RG 1.183 is specified in SRP Section 15.0.3 to be used for new reactors. Not Applicable 1.5 Safety Guide 5 - Assumptions Used for Evaluating the Potential Radiological Conse-quences of a Steam Line Break Accident for Boiling Water Reactors Not Applicable This guidance is only applicable to BWRs. Not Applicable 1.6 Safety Guide 6 - Indepen-dence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems Partially Conforms The onsite electrical AC power systems do not contain Class 1E distribution systems. The EDSS design conforms to the guidance for independence of standby power sources and their distribution systems. 8.3 1.7 Control of Combustible Gas Concentrations in Contain-ment 3 Not Applicable The containment vessel integrity does not rely on combustible gas control systems. Not Applicable 1.8 Qualification and Training of Personnel for Nuclear Power Plants 3 Not Applicable Site-specific programmatic and operational activities are the responsibility of the COL applicant. Not Applicable 1.9 Application and Testing of Safety-Related Diesel Genera-tors in Nuclear Power Plants 4 Not Applicable The NuScale design does not require or include safety-related emergency diesel gen-erators. Not Applicable 1.11 Instrument Lines Penetrating the Primary Reactor Contain-ment 1 Not Applicable No lines penetrate the NPM containment. Not Applicable

NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Tier 2 1.9-7 Draft Revision 2 1.20 Comprehensive Vibration Assessment Program for Reactor Internals During Pre-operational and Initial Startup Testing 3 Partially Conforms The aspects of this RG that mandate start-up tests or inspections for non-prototype reactors are applicable as these tests must be accom-modated in the start-up of each installed NuS-cale Power Module. The remainder is applicable to the COL applications for proto-type reactorsThe first operational NPM is clas-sified as a prototype in accordance with RG 1.20. Thus, the portions of this RG which apply to prototype reactors are applicable to the first operational NPM. After the first NPM is qualified as a valid prototype, subsequent NPMs are classified as non-prototype category I and the non-prototype portions of the RG apply. The identification of departures from RG 1.20 is the responsibility of the COL applicant. 3.9 5.4.1.3 1.21 Measuring, Evaluating, and Reporting Radioactive Mate-rial in Liquid and Gaseous Effluents and Solid Waste 2 Partially Conforms Site-specific, programmatic and operational aspects are the responsibility of the COL appli-cant. 11.5 1.22 Periodic Testing of Protection System Actuation Functions 0 Conforms None. 7.2 1.23 Meteorological Monitoring Programs for Nuclear Power Plant 1 Not Applicable This guidance is the responsibility of the COL applicant. Not Applicable 1.24 Assumptions Used for Evalu-ating the Potential Radiologi-cal Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure 0 Not Applicable Site-specific guidance is the responsibility of the COL applicant. Not Applicable Table 1.9-2: Conformance with Regulatory Guides (Continued) RG Division Title Rev. Conformance Sta-tus Comments Section

RAIO-0518-59932 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com : Affidavit of Zackary W. Rad, AF-0518-59933

AF-0518-59933 NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows: I am the Director, Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I 1. 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. I am knowledgeable of the criteria and procedures used by NuScale in designating 2. 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: The information requested to be withheld reveals distinguishing aspects of a process a. (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. The information requested to be withheld consists of supporting data, including test b. 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. Use by a competitor of the information requested to be withheld would reduce the c. competitor's expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. The information requested to be withheld reveals cost or price information, production d. capabilities, budget levels, or commercial strategies of NuScale. The information requested to be withheld consists of patentable ideas. e. Public disclosure of the information sought to be withheld is likely to cause substantial 3. harm to NuScale's competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the method by which NuScale develops its power module seismic analysis. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment.

AF-0518-59933 The information sought to be withheld is in the enclosed response to NRC Request for 4. Additional Information RAI No. 386, eRAI 9316. 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. The basis for proposing that the information be withheld is that NuScale treats the 5. 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). Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for 6. consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: The information sought to be withheld is owned and has been held in confidence by a. NuScale. The information is of a sort customarily held in confidence by NuScale and, to the best b. 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. The information is being transmitted to and received by the NRC in confidence. c. No public disclosure of the information has been made, and it is not available in public d. 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. Public disclosure of the information is likely to cause substantial harm to the e. competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on 5/10/2018. Zackary W. Rad}}