ML25275A437

Enclosure 1: Crn Site CPA, PSAR Section 1.5, Further Technical Information-Seismic Analysis (Public Version)

ML25275A437
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Site:	05000615
Issue date:	10/01/2025
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
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NNP-25-009
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U.S. Nuclear Regulatory Commission NNP-25-009 October 1, 2025 ENCLOSURE 1 CRN Site CPA, PSAR Section 1.5, Further Technical Information-Seismic Analysis (Public Version)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.5-1 Revision 0 Supplement 1 1.5 Requirements for Further Technical Information This section identifies, describes, and discusses those safety features or components for which further technical information is required in support of the issuance of a construction permit, but which have not been supplied with the CPA.

The following criteria were used to identify activities that require further technical information for Safety Category 1 functions:

Unique design features that support the safety analysis

Adequate margin of conservatism of unique design features that support the safety analysis

Confirmation of design features for extremely low probability events (e.g., Design Extension Conditions (DEC))

For each activity, this section:

1.

Describes the specific technical information that must be obtained to demonstrate acceptable resolution, 2.

Describes how the information will be obtained or cross-reference the PSAR section(s) in which the information will be provided, 3.

Provides a schedule of completion as related to the projected CRN-1 startup date, and 4.

Describes the design alternatives or operational restrictions available if the results do not demonstrate an acceptable resolution.

Design features that are expected to meet established performance criteria, but are subject to on-going evaluations and development of detail design, are not included in this section provided that:

Similar design configurations have been previously approved by the USNRC

Applicable operating experience is judged applicable to CRN-1 conditions

Existence of accepted and applicable codes and standards On-going activities, where applicable, are described in related PSAR sections.

1.5.1 Seismic Analyses A. Description Sections 3.7 and 3.8 describe the seismic and structural design criteria for the BWRX-300 Seismic Category I structures. Appendices 3B through 3H provide a summary of the preliminary analyses that demonstrate how the design of the Seismic Category I structures and associated

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.5-2 Revision 0 Supplement 1 foundation comply with the design criteria. The following Appendices will be provided in a Supplement:

Appendix 3B - Integrated Reactor Building Finite Element Model: Appendix 3B presents an overview of the 3-D FE model used in the structural analyses of the integrated RB and describes the various analysis cases and checks performed to validate the model.

Appendix 3C - Seismic Soil Structure Interaction Analysis Results: Appendix 3C provides key seismic stress demands calculated for key structural members of the integrated RB and assesses the effects of subgrade properties on the seismic response of the soil-structure system by comparing Soil-Structure Interaction (SSI) analysis cases.

Appendix 3D - Interaction Evaluations Results: Appendix 3D provides interaction evaluations results of the Seismic Category II and RW-IIa Power Block structures and foundations with the Seismic Category I structures and foundations under extreme loading.

Appendix 3E - Responses to Static Loads: Appendix 3E presents key design parameters and key results of the 1-g static SSI, static and quasi-static, and thermal load cases considered bounding for the design of the BWRX-300 integrated RB structures to identify critical demands on the structures.

Appendix 3F - Design Details and Evaluation Results for the Containment: Appendix 3F provides bounding estimates of the available margins for the SCCV structure based on results from bounding load combinations selected for the evaluation of the containment and discusses the design of critical SCCV Diaphragm Plate Steel-Plate Composite (DP-SC) connections.

Appendix 3G - Design Details and Evaluation Results for the Containment Internal Structures: Appendix 3G provides bounding estimates of the available margins for the RPV Pedestal structure based on results from bounding load combinations selected for the evaluation of the RB structure.

Appendix 3H - Design Details and Evaluation Results for the Reactor Building Structure:

Appendix 3H provides bounding estimates of the available margins for key structural members of the RB based on results from bounding load combinations selected for the evaluation of the RB structure and discusses the design of critical RB DP-SC connections.

B. Schedule Expected Availability: A Supplement will be provided with the Appendices 3B through 3H in the 4th quarter of FY 2025.

C. Alternatives If the preliminary analyses provided in these appendices do not result in the expected margins to the required acceptance criteria, the structural design features will evolve, as

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.5-3 Revision 0 Supplement 1 required, to ensure all structural and seismic design criteria are satisfied. Any updates will be communicated to the NRC and provided via a planned update during the CPA review period.

Supplement 1, LCR 25-050 1.5.2 10 CFR 50.34(a)(1)(ii)(D) Evaluation A. Description An evaluation assuming a fission product release based on a hypothetical event sequence resulting in core melt will be performed in accordance with 10 CFR 50.34(a)(1)(ii)(D). The hypothetical event sequence and associated fission product release is design-specific and will be analyzed using Modular Accident Analysis Program (MAAP) or suitable alternative as a Design Extension Condition. The ARCON computer program will be used to evaluate the short-term offsite and on-site atmospheric dispersion factors

. The dose consequences will be calculated using RADTRAD and evaluated against dose criteria specified in 10 CFR 50.34(a)(1)(ii)(D) and PDC 19, Control Room.

B. Schedule Expected Availability: Supplement will be provided with the described evaluation within six months of the CPA submittal.

C. Alternatives Additional iteration between design and analysis may be necessary to achieve compliance with acceptance criteria. Any updates will be communicated to the NRC and provided via a planned update during the CPA review period.

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Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-1 Revision 0 Supplement 1 APPENDIX 3B INTEGRATED REACTOR BUILDING FINITE ELEMENT MODEL 3B.1 INTRODUCTION This appendix presents an overview of the 3-D finite element model used in the structural analyses of the integrated Reactor Building and describes the various analysis cases and checks performed to validate the model.

Appendix 3B will be supplied in a a Supplement in the 4th quarter of Government Fiscal Year (FY) 2025.

Supplement 1, LCR 25-050 This appendix presents the Three-Dimensional (3D) Finite Element (FE) models developed for the analysis and design of the integrated Reactor Building (RB) structures which consist of the RB structure outside containment, containment, containment internal structures and their common mat foundation.

Section 3B.1 presents the integrated RB 3D FE model used to perform the various structural analyses discussed in Subsection 3.8.4.1.4. This model is developed using appropriate element types to represent the mass, stiffness, damping characteristics of the integrated RB structures for the analyzed loads being evaluated and resulting stress responses. The model has a sufficiently refined mesh to accurately capture the dynamic response of the structures over the frequency range of interest and to enable accurate calculation of structural stress demands. Mesh refinements in areas of geometric structural discontinuities are used, as applicable.

Table 3B-1 summarizes the integrated RB modeling requirements for the various analysis cases performed. Section 3B.2 presents the 3D FE model used for the seismic Soil-Structure Interaction (SSI) analysis of the integrated RB. Section 3B.3 presents the 3D FE models used for the 1-g static SSI analyses. Section 3B.4 presents the 3D FE models used for static and quasi-static analyses, while Section 3B.5 presents the 3D FE models used for thermal stress analysis. These models are developed per the methodology presented in Subsection 3.7.2.3 and Subsection 3.8.4.1.4.

3B.1 Integrated Reactor Building Finite Element Model 3B.1.1 Model Overview The integrated RB 3D FE model is developed using ANSYS (See Appendix 3I for description) and represents the major structural elements of the integrated RB structures as described in Subsection 3.7.2. The origin of the model is defined at the center of the integrated RB structures with the Z-origin defined at grade (Elevation 0.0 m (0.0 ft)) and the global X-and Y-axes pointed in the North-South, and East-West directions, respectively.

The integrated RB FE model is developed using wall centerlines and approximate centerlines of slabs. The elevation of slabs is adjusted so that the centerline of the slab at grade is at Elevation 0.0 m (0.0 ft). The centerline elevations for the roof slab approximate the weighted average of the slab thickness.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-2 Revision 0 Supplement 1 Figure 3B-1 is a 2/3 slice of the integrated RB FE model showing the major structural components of the model. Containment mechanical and electrical piping penetrations are incorporated in the integrated RB FE model as shown in Figure 3B-2.

The RPV and internals, including fuel, valve mass and supports, are integrated into the integrated RB FE model using a Lumped Mass Stick (LMS) model developed as described in Subsection 3.7.3.3.3 and shown in Figure 3B-3. Subsection 3B.1.4 discusses the connections used to integrate the RPV LMS and the integrated RB models.

3B.1.2 Reactor Building Steel-Plate Composite Structures The integrated RB structures are primarily constructed using Diaphragm Plate Steel-Plate Composite (DP-SC) modules as described in Section 3.8. Geometric details of a typical DP-SC module are illustrated in Figure 3B-4. A DP-SC module system consists of multiple steel components arranged and welded together to form a module. The DP-SC steel faceplates serve as the main reinforcement and permanent formwork for the DP-SC concrete infill. Diaphragm plates with holes allow the flow of concrete inside the module. The diaphragm plates and headed steel studs provide the composite action between the DP-SC steel faceplates and concrete infill, with the diaphragm plates also serving as out-of-plane shear reinforcement for the composite section. The DP-SC modules are spliced together using welded connections to form structural walls, floors, or mat foundation sections as shown in Figure 3B-4. The RB exterior wall and floors, pool walls (excluding fuel pool walls), 36 in thick steam tunnel wing walls, the Steel-Plate Composite Containment Vessel (SCCV) wall and top slab, the RPV pedestal and the common mat foundation are constructed using DP-SC modules.

Conventional steel-plate composite modules, where the two faceplates are connected using discrete tie bars instead of diaphragm plates as is the case for DP-SC modules, are used in the construction of the RB remaining wing walls, partition walls, elevator walls, stair walls and fuel pool walls.

3B.1.3 Model Properties Material properties used in the construction of the integrated RB FE model are described in Section 3.8 and presented in Table 3B-2 and Table 3B-3. Properties assigned to the integrated RB FE model reflect these properties and are computed in accordance with the guidelines of Section 5.0 of NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel (SCCV) and Reactor Building Structural Design, (Reference 3B-1).

Thermal conductivity is determined with ACI/TMS 122R-14, Guide to Thermal Properties of Concrete and Masonry Systems, (Reference 3B-2) and following the guidance of Section 5.5 of NEDC-33926P.

OBE and SSE damping values applicable to the integrated RB structural model are listed in Table 3.7-5.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-3 Revision 0 Supplement 1 3B.1.3.1 Steel-Plate Composite Equivalent Properties The steel-plate composite elements of the integrated RB structures are modeled using elastic shell elements with an equivalent thickness, elastic modulus, and material density calibrated to match the steel-plate composite effective stiffness and mass properties developed using the procedures described in Section 5.5 of NEDC-33926P. The effect of stainless-steel liners used for corrosion or shielding protection is not considered in the calculation of the FE equivalent section properties.

Effective material properties used for the operating conditions represent the Best Estimate (BE) structure stiffness and combine both the steel and concrete stiffnesses as discussed in Section 5.5 of NEDC-33926P. Effective material properties for the fully cracked condition are referred to as the Lower Bound (LB) structure stiffness. Most analyses are performed using the operating/effective (BE) section properties. Thermal accident conditions utilize partially cracked models with section properties that use the operating/effective properties over most of the RB but fully cracked properties over the components subjected to accident conditions. The fully cracked condition is simulated by not considering the contribution of the concrete infill for out-of-plane flexural stiffness and a LB shear stiffness as discussed in Section 5.5 of NEDC-33926P.

Wall curvatures are not considered in the mathematical formulations of steel-plate composite properties but are included in the 3D FE model as discussed in Section 5.12 of NEDC-33926P.

3B.1.4 Steel-Plate Composite Connections The RB FE model utilizes centerline shell element models of steel-plate composite components where adjacent component elements share nodes at their intersecting lines to simulate the steel-plate composite welded connections. This modeling approach effectively suggests rigid connection behavior, with no relative deformations between components.

The below grade slabs and wing walls are connected to the SCCV using six independent springs at each intersecting node to represent the stiffness in each of the deformation degree of freedom.

These springs are assigned a large stiffness value to simulate rigid connection behavior.

The centerlines of shell elements representing slabs and walls are connected in the RB model.

The RB model includes an enhanced feature within the below grade slabs that incorporates rigid shell elements at the slabs inner and outer radial edges to account for the rigidity that exists within the slab-to-wall intersection. The slab mesh radial dimension is set equal to half the wall thickness and the elements assigned an elastic modulus equal to 20 times the slab modulus to provide the rigid offset.

3B.1.5 Reactor Pressure Vessel and Containment Equipment and Piping Support Structure Connections The BWRX-300 Reactor Pressure Vessel (RPV) is supported vertically and laterally by a steel skirt anchored using anchor bolts to structural steel that lines the top of the pedestal as seen in Figure 3.8-1. Top and bottom stabilizers (shown in Figure 3.8-1) provide additional lateral support to the RPV. The Control Rod Drive Housing (CRDH) located below the RPV (see Figure 3.8-1) is

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-4 Revision 0 Supplement 1 also supported by lateral supports. The RPV is free for radial and vertical movement at the stabilizers and CRDH supports. The skirt, bottom stabilizers, and CRDH lateral support transfer lateral load directly to the pedestal. The upper stabilizers transfer lateral load to the Containment Equipment and Piping Support Structure (CEPSS).

The RPV LMS model shown in Figure 3B-1 and Figure 3B-3 is developed to represent the dynamic mechanical behavior of the BWRX-300 RPV, reactor internals, fuel, valve mass, and its supports stiffness. During its incorporation into the integrated RB FE model, the LMS model was adapted to ensure a proper transfer of RPV loads to the pedestal and CEPSS framing shown in Figure 3B-1 and Figure 3B-3.

The CEPSS consists of a 3D space frame that supports three platform levels. Each platform includes radial and circumferential beams along with infill beams that will support the platform grating and large equipment and piping supports. The CEPSS framing includes vertical bracing members around the CEPSS outer perimeter (near SCCV wall) and inner perimeter (above the pedestal). Horizontal bracing members are also included integral to the platform framing which are critical for lateral force transfer to the SCCV wall. The CEPSS framing is vertically and laterally supported by the pedestal and tangentially restrained by the SCCV wall through shear lugs that allow the vertical and radial relative expansion between the CEPSS and SCCV.

3B.1.6 Modeling of Openings and Penetrations The model mesh size meets the ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, (Reference 3B-3) modeling requirements for the purposes of element strength evaluation including modeling in the connection regions and around penetrations (mesh size < tsc, where tsc is the depth of the steel-plate composite module as indicated in Figure 3B-4).

Where known large penetrations (> tsc/2) exist, the opening is included in the model using the actual or approximate opening size. Where known small penetrations (< tsc/2) exist, the opening is not explicitly modeled and evaluation around the small openings depends on whether the penetration is detailed as free edge or fully developed.

3B.1.7 Static and Dynamic Mass Properties The integrated RB model is analyzed for the design loads described in Section 3.8. These loads are assigned to the integrated RB model as static and dynamic masses.

3B.1.7.1 Equipment and Non-Structural Mass The dead inertial mass assigned to the integrated RB model includes the self-weight and supplemental weight added to represent equipment and non-structural dead loads not represented in the model. Static dead load analysis is performed by applying a 1-g acceleration to the SSI model including dead load masses. The dead load inertial masses also contribute to the seismic inertial loads and remain in the integrated RB seismic SSI analyses.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-5 Revision 0 Supplement 1 The non-structural dead load mass assigned to the integrated RB model includes the mass of staircases, partition walls, the basemat 2 ft thick normal weight concrete topping slab, the pedestal refractory material (weighing approximately 820 psf), and the containment and CEPSS platforms. The assigned mass of the two platforms inside containment is equivalent to 65 psf.

The CEPSS top and middle platforms are assigned 75 psf for equipment dead load and 16 psf for light weight steel grating. The CEPSS lower platform is assigned 75 psf for equipment dead load and 32 psf for heavy steel grating.

In areas where the concentrated equipment loads are larger than 100 psf, a uniform mass is added to the shell elements by increasing the material density. Equipment masses are accounted for by increasing floor shell element density. Concentrated mechanical and electrical equipment load mass is assigned in two ways in the model, either by increasing the density of the floor shell elements, or by assigning the mass to a set of nodes in the model.

Outside of the heavy equipment areas, a uniformly distributed load of 100 psf is applied to the floors of the Integrated RB to simulate the additional static mass from mechanical, electrical, and miscellaneous equipment supported from the floors.

The polar crane composed is of an overhead bridge of two deep girders supporting a trolley, with the polar cranes parked position aligned with the Y axis of the RB. The seismic analysis model includes vertical and horizontal mass at each girder end representing the crane girders. The trolley weight is also included as vertical and horizontal masses with the trolley parked at the end of the crane in the +Y direction. Vertical mass representing the full 125-ton payload is also included for the seismic model and positioned at the end of the crane in the +Y direction. Due to the pendulum effect of the suspended payload, the horizontal inertia mass of the payload is not included in the seismic model. The 1-g analysis model only includes the vertical mass of the crane girders and trolley (parked at the end of the crane in the +Y direction) as this analysis model is intended to include the structural dead loads. As indicated in Table 3B-1, the crane payloads and operational loads are considered in a separate load case defined in Subsection 3E.1.8.

3B.1.7.2 Equivalent Live and Snow Loads A 100 psf live load is considered for all floors except for the refueling floor where the live load is 200 psf. The seismic model includes fifty percent of the live load on floor areas without heavy equipment and twenty-five percent of snow loads are modeled.

No equivalent live load mass is considered on the CEPSS for the seismic model as it is within the containment and no live load is present during normal operating conditions.

3B.1.7.3 Hydrostatic and Hydrodynamic Mass The hydrostatic and hydrodynamic loads applied on the integrated RB 3D FE model are computed using best estimate water levels in pools under normal operating conditions.

Methodology used to account for hydrodynamic effects on the integrated RB pools is discussed in Subsection 3.7.2.3.1.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-6 Revision 0 Supplement 1 Hydrostatic loads on pool floors are captured by modeling the fluid volume in each of the pools as a vertical only mass during 1-g static SSI analyses. Hydrostatic loads on pool walls are analyzed as a pressure load during the static structural analysis as demonstrated in Subsection 3E.3.2.

The vertical fluid volume mass on pool floors also acts as the seismic inertial mass in the vertical direction during seismic analysis. The hydrodynamic impulsive loads on pool walls are captured during seismic response analysis by assigning masses to the pool walls consistent with the impulsive pressure distribution. Hydrodynamic convective (sloshing) loads and the breathing mode lateral pressures on pool walls are analyzed as a pressure load during the static structural analysis. The breathing mode lateral pressures are the result of vertical seismic excitation on the pools causing an increased vertical pressure and lateral pressures with distributions matching the hydrostatic pressures but with magnitudes based on the seismic response at the pool levels.

3B.2 Seismic Structure-Soil-Structure Interaction Analysis Models As the integrated RB is deeply embedded, the interaction of the structure with the surrounding subgrade is important for its structural integrity and its response under static and dynamic loads.

In accordance with the guidance of Section 5.1 of NEDO-33914-A, BWRX-300 Advanced Civil Construction and Design Approach, (Reference 3B-4), the one-step approach, as defined in Section 3.1.2 of ASCE/SEI 4-16, Seismic Design of Safety-Related Nuclear Structures, (Reference 3B-5), is implemented for the design of the BWRX-300 integrated RB structures. To adequately account for the SSI effects, a 3D FE model that includes FE representations of the integrated RB (see Section 3B.1 and Figure 3B-1) and the surrounding subgrade is developed as described in Subsection 3.7.2. The design basis analysis case model for the RB seismic SSI analysis, shown in Figure 3B-5, includes the explicit representation of certain construction elements in the Clinch River Nuclear (CRN) excavation plan which are illustrated in Figure 3B-6.

These include the mud mat under the RB mat foundation shown in red in Figure 3B-5 and Figure 3B-6, and the concrete fill in the annulus and shelf shown in purple and red in the same figures. The model also includes solid elements, shown in blue in Figure 3B-5 and Figure 3B-6, that may be assigned material properties corresponding to soil or simulating the temporary excavation support to assess their effects on the RB seismic response as discussed in Subsection 3.7.2.3 and Subsection 3.7.2.9.2. This model is paired with the range of in-situ as-built subgrade conditions discussed in Subsection 3.7.1.3 to perform the Case 1 seismic SSI analysis documented in Appendix 3C.

Coarse surface mounted FE models representing the global dynamic properties and weight of the surrounding Power Block structures and foundations are coupled in ANSYS with the Integrated RB and subgrade model shown in Figure 3B-5 to capture the Structure-Soil-Structure Interaction (SSSI) effects as discussed in Subsection 3.7.2.4. The coarse Power Block models are inserted into the integrated RB model with the centerline of their mat foundations set a grade elevation (Elevation 0.0 m (0.0 ft)). This combined SSI model, shown in Figure 3B-7, is paired with a second case of as-built subgrade conditions where the in-situ soil and weathered rock are excavated down to competent rock and replaced with engineered fill up to plant grade Elevation 814.5 ft to provide adequate support for supporting the adjacent Power Block structures. The results of this Case 2 seismic SSI analysis are also summarized in Appendix 3C.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-7 Revision 0 Supplement 1 The Turbine Building (TB), one of the Power Block structures included in the combined SSI model, is divided into three separate structural systems consisting of the TB shell structure, radiological Shield Wall Area, and the turbine generator pedestal. The turbine generator pedestal including foundation is structurally isolated from both the TB shell and the shielding walls for vibration control. Figure 3B-8 shows the LMS model used to represent the turbine generator pedestal structure, and the isolated FE model used to model its isolated foundation in the combined seismic SSI analysis model.

The Case 1 and Case 2 seismic SSI models, also referred to as the dynamic SSI models, are assigned BE structure stiffnesses with seismic and dead load inertial masses, and lower OBE-level damping properties. The models are developed in ANSYS and then translated into ACS SASSI (see Appendix 3I) format.

3B.2.1 Modeling of Soil-Structure Interfaces The soil and soil-structure interfaces of the dynamic SSI analysis models in SASSI consist of the following:

Far-field subgrade model representing the properties of the subgrade materials that are approximated as continuum infinite horizontal layers resting on surface of uniform half-space.

Near-field subgrade model representing the properties of subgrade materials supporting the adjacent Power Block foundations and around the RB below grade exterior wall.

Excavated soil volume representing the properties of the volume of far-field subgrade replaced by the RB and adjacent near-field volume elements.

Linear spring elements to connect the near-field subgrade elements to structural elements, which are assigned stiffness properties that can be adjusted to simulate different interface conditions. The excavated volume and the near-field volumes are also connected with linear spring elements at the near-field (backfill) perimeter.

The soil elements used in dynamic and subgrade impedance calculations use a consistent layer thickness governed by the dynamic response which is sufficiently refined to transmit the entire frequency range of interest. The minimum layer thickness is approximately double the shell element mesh size.

3B.2.1.1 Far-Field Subgrade The far-field subgrade is represented in ACS SASSI by half-space continuum of horizontally infinite layers with uniform linear elastic properties resting on uniform elastic half-space. A viscous lower boundary of the subgrade model is established at a depth of 482 ft meeting the regulatory guidance of Subsection II.4 of SRP 3.7.2 as discussed in Subsection 3.7.2.3.1.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-8 Revision 0 Supplement 1 3B.2.1.2 Near-Field Subgrade Volume Figure 3B-9 presents the single layer of near-field solid elements placed below the BWRX-300 Power Block foundations, which are assigned subgrade material properties, and the solid FEs around the below grade portion of the RB exterior wall that represent the properties of the concrete backfill in the two seismic SSI analysis cases.

At the excavated volume perimeter, the FE of the excavated soil volume (see Subsection 3B.2.1.3) and near-field volume share the same nodes. The mesh of these two subgrade volumes aligns with every other element of the RB exterior wall and mat foundation mesh. Having the FE mesh of the near-field volume and excavated volume being twice the size of the refined one-step approach RB structural model helps reduce the ACS SASSI dynamic soil impedance calculations to reasonable durations. The mesh of the near-field volume also aligns with the mesh of the surface mounted Power Block foundations.

3B.2.1.3 Far-Field Subgrade Excavated Volume The excavated volume model (see Figure 3B-10) consists of 3D solid FEs representing the part of the horizontally layered subgrade displaced by the near-field volume (see Subsection 3B.2.1.2) and the embedded portion of the RB structure. The mesh of the excavated subgrade volume aligns with the layering of the far-field subgrade profiles and the near-field subgrade volume shown in Figure 3B-9. The meshes of the excavated and near-field volumes are aligned at the perimeter, where the excavated volume and near-field volume solid elements share the same nodes.

3B.2.1.4 Near-Field - Structure Interfaces Three linear spring elements are assigned to nodes between the near-field volume and the FE models of the RB and adjacent Power Block structures as well at the nodes of the RB exterior wall to represent the interaction mechanisms at the structure and soil interfaces. These spring elements are assigned to approximately every other node on the RB exterior wall and mat foundation mesh.

For seismic analysis, interface springs in the dynamic SSI models are assigned high stiffnesses to simulate bonded conditions between the near-field volume solid elements and the structural shell elements and the near-field elements and the excavated volume.

3B.3 1-g Static Soil-Structure Interaction Analysis Model The 1-g static SSI analysis provides design demands for gravity inertia loads and static earth pressure loads including surcharge loads from the foundations of adjacent Power Block buildings.

Two sets of 1-g static SSI analysis are performed to evaluate the unfactored and factored gravity demands on the integrated RB (see Table 3B-1) requiring the development of two 1-g static SSI analysis models. Factored gravity demands are obtained from load combinations with different load factors assigned to dead and soil pressure loads (1.2D+1.6H or D+1.3H).

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-9 Revision 0 Supplement 1 The two 1-g static SSI models have the same configuration and FE attributes as the combined dynamic model developed to capture the SSSI effects described in Section 3B.2. The structural members of the 1-g static SSI models are assigned the same BE effective stiffness properties as the dynamic models. Differences between the static and dynamic models consist of the following adjustments made to the unit weight properties of shell and lumped mass elements of the 1-g SSI models:

The removal of the live, roof snow and crane mass from the 1-g static SSI Models. The fluid horizontal dynamic masses are not removed from the models as these are not activated in 1-g gravity analysis.

The total mass of adjacent Power Block structures is increased by 10% in the unfactored and factored 1-g SSI analysis models for added safety margin. The factored 1-g SSI analysis also uses unit weight properties of the subgrade and surrounding structures increased by an additional 33%.

Per recommendations of Section 5.1.2 of NEDO-33914-A, stiffness properties assigned to the spring elements in the 1-g static SSI models provide SSI interface conditions that emphasize the lateral earth pressure loads on the RB exterior wall below grade. An Upper Bound (UB) stiffness is assigned to the spring elements in the direction normal to the RB exterior wall and mat foundation to simulate bonded conditions. The friction at the RB exterior wall and mat foundation is not considered by assigning very low stiffness properties to the contact springs in the vertical and tangential direction.

3B.4 Static and Quasi-Static Analysis Models The static and quasi-static analyses provide design demands for static normal operating and accident conditions loads, excluding thermal loads, not captured in the 1-g static SSI analysis.

For more information on load cases considered in the analyses, refer to Subsection 3.8.4.1.4 and Table 3B-1.

As indicated in Table 3B-1, static and quasi-static analyses are performed on standalone models of the integrated RB with either BE structure stiffness properties for normal operating and environmental loads or partially cracked conditions consisting of effective stiffness properties for all components except those subjected to temperature demands that create fully cracked concrete conditions.

Static and Quasi-static analyses are performed using the subgrade stiffness sub-structuring method as discussed in Subsection 3.8.4.1.4 in which the subgrade represented by solid FEs (see Figure 3B-11) is condensed into a single matrix element, called a super-element. This super-element represents the SSI stiffness at nodes coincident with every other node of the mat foundation and every node at the outside of the near-field volume. For static and quasi-static analyses, the super-element is assigned LB subgrade stiffness properties to emphasize the deformations at the subgrade-structure interfaces. This results in conservative estimate of the member force and moment demands on the RB integrated structure.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-10 Revision 0 Supplement 1 3B.5 Thermal Stress Analysis Model Thermal analyses provide stress demands due to normal operating and accident load conditions as indicated in Table 3B-1.

Thermal stress analyses are performed on standalone models of the integrated RB that have the same configuration and FE attributes as the models developed for static and quasi-static analyses discussed in Section 3B.4. The standalone integrated RB model used for normal operating thermal analysis is developed with BE structure stiffness properties. The standalone integrated RB model used for the accident thermal analysis is developed with lower structure stiffness properties (fully cracked or lower bound section properties) assigned to walls and slabs within DBA regions affected by DBA thermal loads (e.g., the exterior pool walls, SCCV wall, RPV pedestal, inner mat foundation and SCCV top slab), and BE structure stiffness properties elsewhere.

In addition to the modeling requirements of Table 3B-1, the following modifications are also included in the FE models used for thermal stress analyses:

The SCCV and RPV pedestal rigid beams used to support the dynamic analysis are removed or modified to have a very low elastic modulus and thermal coefficient to match the steel-plate composite they are embedded into.

The RPV LMS model developed for the dynamic analysis is also removed from the models to obtain accurate thermal results. The removal of the RPV LMS model is acceptable as the RPV is designed to allow nearly unrestrained thermal growth through the stabilizers and RPV skirt.

Crane rigid support offsets are removed to eliminate artificially high thermal stresses.

Various conditions for normal operating and DBA thermal stress analyses are considered and use super-elements assigned LB and UB subgrade stiffness properties to obtain bounding internal forces for the design of the integrated RB structures.

3B.6 Validation of Finite Element Models This section documents the verification of the integrated RB ANSYS seismic analysis models and their translation from ANSYS to SASSI.

The verification process includes detailed review of the assigned finite elements to create the models structural components, structural connectivity, and the non-structural masses along with static, modal, and harmonic analyses to confirm model static and dynamic behavior. The analysis models include the masses, structural stiffness, and damping appropriate for seismic analyses.

The analysis models are translated to SASSI for use in the SSI analyses documented in Appendix 3C. These models are additionally used as a basis for static and thermal structural analysis in ANSYS with some adaptations as described in Section 3B.3 through Section 3B.5.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-11 Revision 0 Supplement 1 In addition to a detailed review of the RB FE model, the model validation includes the following steps:

Modal analysis of the RB fixed base model

1-g analysis of the RB fixed base model in X-, Y-, and Z-direction

Modal analysis of the isolated RPV LMS model for verification of adaptation into the integrated RB model

Modal analysis of the Power Block structures for verification of adaptation into the integrated RB model

Transfer function comparison between ANSYS and SASSI RB fixed base models for verification of translation of the integrated RB model to SASSI The verification steps discussed in Subsection 3B.6.1 through Subsection 3B.6.5 demonstrate that the integrated RB ANSYS structural models adequately represent the RB static and dynamic structural behavior and confirm the accurate translation of the model into SASSI.

3B.6.1 Fixed Base Modal Analysis of Reactor Building The objective of the modal analysis is to understand the key dynamic properties of the model.

The Integrated RB significant mode frequencies, and mass participation ratios are presented in Table 3B-4. These modal frequencies agree reasonably well with simplified hand calculations which provides confidence in the model.

The integrated RB refined mesh ensures an accurate representation of the dynamic response and accurate computations of structural demands as discussed in the introductory paragraph of this Appendix. In addition, modeling equipment and non-structural masses as a combination of distributed and nodal mass as discussed in Subsection 3B.1.7.1 captures the participation of all mass and ensures an adequate representation of the dynamic modes for the structures.

3B.6.2 Fixed Base 1-g Analysis of Reactor Building A 1-g acceleration analysis is performed for the fixed base RB model in each of the three orthogonal directions to assess the total seismic mass and to examine the deformed shape and ensure component connectivity in all three directions.

The total seismic mass/weight for the fixed base RB model is summarized in Table 3B-5. The deformed shapes of the RB for the 1-g X, 1-g Y and 1-g Z analyses are shown in Figure 3B-12 through Figure 3B-14. The deformed shapes shown in Figure 3B-12 through Figure 3B-14 confirm that element connectivity is as expected between the various structural components, including the steel-plate composite components, the RPV LMS model and the CEPSS.

3B.6.3 Validation of Incorporation of Reactor Pressure Vessel Lumped Mass Stick Model The incorporation of the RPV LMS model into the integrated RB model is validated by comparing the dynamic characteristics of the RPV LMS model within the integrated RB model with benchmark results of the eigenvalue analysis of the standalone RPV LMS model. To perform this

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-12 Revision 0 Supplement 1 comparison, the RPV LMS model is isolated in the global RB FE model by applying fixed boundary conditions to all nodes up to LMS model connection points (i.e., skirt to the top of pedestal, top and bottom stabilizers).

Figure 3B-15 compares the modal properties (frequencies and cumulative masses) between the RPV LMS model before and after integration into the integrated RB model. As demonstrated by Figure 3B-15, the dynamic coupled response between the integrated RPV LMS model and the RB is adequately captured.

3B.6.4 Validation of Dynamic Models of Adjacent Power Block Structures The coarse FE models of the adjacent Power Block structures that are coupled with the integrated RB are required to represent their global dynamic properties for use in seismic SSI analyses.

To validate the translation of each Power Block building into the combined seismic SSI analysis model, a comparison of the dynamic frequencies of the primary modes of vibration in each direction between the provided coarse mesh models and each power block building isolated in the global structural model is performed. The isolation of the global structural models is achieved by fixing the degrees of freedom at each building foundation and the degrees of freedom associated with the structures.

Validation results demonstrate that the dynamic characteristics of the coarse dynamic models match reasonably well in total mass and dominant dynamic response with those from the refined models with the largest frequency difference being 1.7% for the Turbine Building Y-direction response. Therefore, the coarse FE models of the adjacent Power Block structures provide an adequate representation of the global dynamic properties for the SSI analyses of the integrated RB.

3B.6.5 SASSI Model Translation Verification To validate the translation of the integrated RB model from the ANSYS to the SASSI, the total acceleration transfer functions obtained from the ANSYS fixed based RB and the SASSI fixed base model are compared.

The transfer functions are defined as the ratio of the acceleration at a particular point in the structure to the amplitude of the input acceleration at the base. Harmonic analysis is performed in ANSYS to determine the transfer function while SASSI utilizes transfer functions as a key calculated quantity in its response calculation approach.

Figure 3B-16 and Figure 3B-17 compare the two pairs of transfer functions for the center of the RB roof and middle of RPV fuel. As shown in Figure 3B-16 and Figure 3B-17, the transfer function peaks occur at frequencies consistent with the primary mode frequencies obtained from modal analysis, thus confirming the translation of each model from ANSYS to SASSI.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-13 Revision 0 Supplement 1 3B.6.5.1 ANSYS Fixed Base Harmonic Analysis The RB model used in the ANSYS fixed base harmonic analysis is the model shown in Figure 3B-5 excluding the Power Block backfill layer. The harmonic analysis model intentionally keeps the RB backfill, shelf, and additional solid elements along with the inter-connecting spring elements to ensure proper translation of the finite elements up to the connection with the soil layering in SASSI.

Transfer functions are obtained at key locations, such as the pedestal, RPV, CEPSS, and fuel, to ensure the translation can capture the global structural behavior of the integrated RB structures.

A full integration harmonic analysis is then performed, where a harmonic 1-g acceleration input is used to excite the model with frequencies varying from 0.1 to 20 Hz for horizontal input and 0.1 to 40 Hz for vertical input, at increments of 0.1 Hz. The results of harmonic analysis include the amplitude of the steady-state relative displacement and response phase-angle.

3B.6.5.2 SASSI Fixed Base Frequency Domain Analysis SASSI analyses are performed on FE models of the integrated RB mounted on a series of rigid soil layers. To simulate fixed base conditions, the following properties are assigned to the soil layers beneath the integrated RB common mat foundation:

Shear wave velocity (VS) of 32,808 ft/s

Compression Wave Velocity (VP) 65,617 ft/s

Damping ratio of 1%

SASSI analyses of the fixed base models are performed for analysis frequencies ranging from 0.1 to 70 Hz to calculate acceleration transfer function amplitudes for the responses in the three orthogonal directions at the same key nodal locations.

3B.7 References 3B-1 NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel (SCCV) and Reactor Building Structural Design, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, December 2024.

3B-2 ACI/TMS 122R-14, Guide to Thermal Properties of Concrete and Masonry Systems, American Concrete Institute, 2014.

3B-3 ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, American National Standards Institute/American Institute of Steel Construction, 2018.

3B-4 NEDO-33914-A, BWRX-300 Advanced Civil Construction and Design Approach, GE-Hitachi Nuclear Energy Americas, LLC, Revision 2, June 2022.

3B-5 ASCE/SEI 4-16, Seismic Design of Safety-Related Nuclear Structures, American Society of Civil Engineers, 2017.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-14 Revision 0 Supplement 1 Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 1 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model Seismic Inertial, earth

pressure, hydrodynamic:

impulsive pressure on pool walls and slabs E

Seismic SSI (ACS SASSI)

Effective BE BE strain compatible Rigid in all three directions Rigid in all three directions Standalone (for SSI) and Coupled dynamic models of RB and adjacent Power Block structures (for SSSI effects)

LB strain compatible UB strain compatible Unfactored dead load, vertical hydrostatic load, and lateral static earth pressure including adjacent building surcharge D, Fv, H 1-gstatic SSI (ANSYS)

Effective BE UB unit weight Rigid in radial (bearing) directions. Soft in vertical/tangential (frictional) directions Rigid in vertical direction only Coupled static models of the RB and adjacent Power Block structures UB Poissons ratio Factored dead load, vertical hydrostatic load and lateral static earth pressure including adjacent building surcharge LB Youngs modulus

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-15 Revision 0 Supplement 1 Live load on floors L

Static subgrade stiffness sub-structuring analysis (ANSYS)

Effective BE LB static massless LB Poissons ratio LB Youngs modulus Rigid in radial (bearing) directions. Soft in vertical/tangential (frictional) directions Rigid in vertical direction only Standalone RB model Crane C

Hydrostatic and hydrodynamic breathing mode pressures on pool walls Fh or Eb Wind load on above grade RB exterior wall and roof W

Tornado wind load above grade RB exterior wall and roof Wt Containment accident flooding Ha Subgrade slab flooding Hg Normal operating and Structural Integrity Test uniform containment pressure P0, Pop, Pt Accident uniform containment pressure Pa LB for walls/

slabs within DBA region affected by DBA thermal loads and effective BE elsewhere Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 2 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-16 Revision 0 Supplement 1 Accident seismic torsion Et Static subgrade stiffness sub-structuring analysis (ANSYS)

Effective BE LB static massless LB Poissons ratio LB Youngs modulus Soft in all three directions Rigid in all three directions Standalone RB model Seismic hydrodynamic sloshing load on pool walls Es Normal snow load on roof S

Extreme precipitation (snow) pressure on roof Sx Groundwater hydrostatic pressure on mat foundation (Hb) and RB exterior wall (Hg) below groundwater level Hg, Hb Rigid in vertical direction only except at mat foundation center rigid in all three directions Normal operating and DBA nozzle, equipment and piping reaction loads Ro, Ra Local loads on containment during DBA Rr, Rb Additional lateral rock pressure on RB exterior wall Hr Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 3 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-17 Revision 0 Supplement 1 Normal operating thermal loads TC1 -

Summer Thermal stress subgrade stiffness sub-structuring analysis (ANSYS)

Effective BE UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions Rigid in all three directions Standalone RB model TC2 -

Summer LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions TC3 - Winter UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions TC4 - Winter LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 4 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-18 Revision 0 Supplement 1 Design basis accident inside containment -

DBA-1 TC5 -

Summer Thermal stress subgrade stiffness sub-structuring analysis (ANSYS)

Effective except SCCV and RPV pedestal use LB UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions Rigid in all three directions Standalone RB model TC6 -

Summer LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions TC7 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> duration) and TC9 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration) -

Winter UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions TC8 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC10 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions Design basis accident inside steam tunnel -

DBA-2 TC11 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC13 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter Thermal stress subgrade stiffness sub-structuring analysis (ANSYS)

Effective except Steam Tunnel uses LB UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions Rigid in all three directions Standalone RB model TC12 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC14 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 5 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-19 Revision 0 Supplement 1 Design basis accident in isolation condenser pools and/or reactor cavity and equipment pools - DBA-3 TC15 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC17 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter Thermal stress subgrade stiffness sub-structuring analysis (ANSYS)

Effective except pool walls/slabs use LB UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions Rigid in all three directions Standalone RB model TC16 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC18 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions Design basis accident in fuel pool - DBA-4 TC19 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC21 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter Thermal stress subgrade stiffness sub-structuring analysis (ANSYS Effective except Fuel Pool walls/

slabs use LB UB static massless UB Poissons ratio UB Youngs modulus Rigid in all three directions Rigid in all three directions Standalone RB model TC20 (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) and TC22 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) -

Winter LB static massless LB Poissons ratio LB Youngs modulus Rigid in normal directions. Soft in the other two directions Table 3B-1 Integrated RB Analyses Cases and Model Requirements (Sheet 6 of 6)

Loading Type Applicable Design Load(s)

Analysis Method (Computer Program)

Structural Stiffness Subgrade Properties Embedded RB Exterior Wall Contact Springs Mat Foundation Contact Springs Structural Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-20 Revision 0 Supplement 1 Table 3B-2 Steel-Plate Composite Material Properties Material Properties Value in Imperial Units Steel Unit weight 490 Ib/ft3 Concrete unit weight Normal Density Concrete 145 Ib/ft3 High Density Concrete 219 Ib/ft3 Compressive strength of concrete 8 ksi 8

ksi Steel Yield Stress grade 50 50 ksi Steel Yield Stress grade 60 60 ksi Steel Yield Stress grade 65 65 ksi Concrete Elastic modulus Normal Density Concrete where is in ksi and in lb/ft3 High Density Concrete

where, is in ksi, is in psi and in lb/ft3 Steel Elastic modulus 29,000 ksi Concrete Poisson's ratio 0.17 unitless Steel Poisson's ratio 0.3 unitless Thermal expansion coefficient 5.5 10-6

/°F Thermal conductivity coefficient in lb/ft3 Btu*in./(h*ft2*°F)

Ec 1.5 c

f c

=

f c c

Ec 234 fc 3

c 144

1.17

=

Ec f c c

kc 0.5e 0.02c

=

c

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-21 Revision 0 Supplement 1 Table 3B-3 Structural Steel Properties Engineering Properties Value Unit weight 490 Ibf/ft3 Young's modulus 29,000 ksi Poisson's ratio 0.3 Thermal expansion coefficient 7.8 x 10-6/ °F

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-22 Revision 0 Supplement 1 Table 3B-4 Significant Modes of the Fixed Base Integrated Reactor Building Model Mode Frequency (Hz)

X Y

Z ROTX ROTY ROTZ Mode Shape Description Effective Mass/Total Mass (%)

1 2.42 40 23 0

4 8

0 Primary X mode, RB, SCCV, CEPSS 2

2.47 23 41 0

7 4

0 Primary Y mode, RB, SCCV, CEPSS 5

5.23 0

0 0

0 0

78 Torsion Mode 6

6.83 0

0 19 0

0 0

RB Roof vertical mode 7

7.15 5

8 0

17 10 0

Y second mode, and Pedestal Y 8

7.19 9

5 0

10 17 0

X second mode, and Pedestal X 13 9.03 0

0 50 0

0 0

Primary Z mode, RB, and SCCV 23 13.01 1

3 0

6 3

0 Y third mode, and Pedestal Y 24 13.15 2

1 0

4 5

0 X third mode, and Pedestal X

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-23 Revision 0 Supplement 1 Table 3B-5 Static Analysis of Fixed Base Integrated Reactor Building Model Direction Seismic Weight (kip)

Max Displacement at Roof (in)

X-direction 1.71E5 2.56 Y-direction 1.73E5 2.49 Z-direction 1.72E5 0.40

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-24 Revision 0 Supplement 1 Figure 3B-1 Overview of Integrated RB Finite Element Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-25 Revision 0 Supplement 1 Figure 3B-2 Steel-Plate Composite Containment Vessel Structural Model with Penetrations

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-26 Revision 0 Supplement 1 Figure 3B-3 Lumped Mass Stick Model of Reactor Pressure Vessel and Containment Equipment and Piping Support Structure

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-27 Revision 0 Supplement 1 Figure 3B-4 Typical Diaphragm Plate Steel-Plate Composite Section and Module System

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-28 Revision 0 Supplement 1 Figure 3B-5 Reactor Building Standalone Soil-Structure Interaction Model (for Case 1)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-29 Revision 0 Supplement 1 Figure 3B-6 Backfill, Shelf and Additional Solid Elements Added to Reactor Building Design Basis Seismic Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-30 Revision 0 Supplement 1 Figure 3B-7 Combined Seismic Soil-Structure Interaction Analysis Model with Clinch River Nuclear Excavation Layout (For Case 2)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-31 Revision 0 Supplement 1 Figure 3B-8 Turbine Generator Pedestal and Isolated Foundation at Grade

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-32 Revision 0 Supplement 1 Figure 3B-9 Reactor Building and Power Block Near-Field Subgrade Volume Elements

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-33 Revision 0 Supplement 1 Figure 3B-10 Reactor Building and Power Block Excavated Subgrade Volume Element Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-34 Revision 0 Supplement 1 Figure 3B-11 Finite Element Model for Calculation of Subgrade Impedance

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-35 Revision 0 Supplement 1 Figure 3B-12 1-g X Primary Deformed Shape (m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-36 Revision 0 Supplement 1 Figure 3B-13 1-g Y Primary Deformed Shape (m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-37 Revision 0 Supplement 1 Figure 3B-14 1-g Z Primary Deformed Shape (m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-38 Revision 0 Supplement 1 Figure 3B-15 Cumulative Mass Participation Ratio for the Lumped Mass Stick Model of the Reactor Pressure Vessel

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-39 Revision 0 Supplement 1 Figure 3B-16 Total Acceleration Transfer Functions at Center of Reactor Building Roof

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3B-40 Revision 0 Supplement 1 Figure 3B-17 Total Acceleration Transfer Functions at Middle of Fuel

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-1 Revision 0 Supplement 1 APPENDIX 3C SEISMIC SOIL STRUCTURE INTERACTION ANALYSIS RESULTS 3C.1 Introduction This appendix provides key seismic stress demands calculated for key structural members of the integrated Reactor Building and assesses the effects of subgrade properties on the seismic response of the soil structure system by comparing Soil Structure Interaction (SSI) analysis cases.

Appendix 3C will be supplied in a Supplement in the 4th quarter of FY 2025.

Supplement 1, LCR 25-050 This appendix documents the results of the seismic Soil-Structure Interaction (SSI) analysis of the integrated Reactor Building (RB).

3C.1 Seismic SSI Analysis Methodology Seismic demands for the design of the Seismic Category I SSCs at the Clinch River Nuclear (CRN) Site are obtained from the SSI analysis of the Three-Dimensional (3D) Finite Element (FE) models discussed in Section 3B.2 evaluated for two free-field site property conditions.

Case 1 considers the in-situ soils to be left intact and engineered backfill is added to raise the plant finished grade elevation to 814.5 ft, while Case 2 considers the condition where the residual soil and weathered rock are removed down to competent rock and replaced with engineered backfill up to the plant finished grade. As stated in Section 3B.2, the Case 1 profile is paired with the standalone RB analysis model (excluding other Power Block structures), while the Case 2 profile is paired with the full FE model combining the RB with the adjacent Power Block structures. The subgrade parameters are developed to conservatively cover the variabilities and address uncertainties in the geotechnical properties of the subsurface materials and seismicity of the CRN Site. The consideration of Cases 1 and 2 provides additional margin in the estimated seismic demands for the design of the Seismic Category I structures in the absence of caissons in the analyzed models.

The Case 1 and Case 2 seismic SSI analyses are performed using the one-step analysis approach discussed in Subsection 3.7.2.1 which allows seismic stress demands to be obtained directly from the results of the SSI analyses. In both cases, the control motion is applied to the model at the bottom of the RB with the vertical propagation of horizontal seismic motions through the subgrade considered to be controlled by the shear-wave velocities of the soil/rock layers, and the propagation of vertical motions considered to be controlled by compression-wave velocities.

As stated in Section 3B.2, the dynamic SSI FE models are assigned structural material damping ratios corresponding to Response Level 1 per Table 3-1 of ASCE/SEI 43-19, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, (Reference 3C-1) for steel-plate composite elements (3%) and metal structures (4%) in lieu of the larger damping ratios corresponding to Response Level 2 as permissible per U.S. Nuclear Regulatory Commission (USNRC) Regulatory Guide (RG) 1.61, Damping Values for Seismic Design of Nuclear Power Plants, (Reference 3C-2).

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-2 Revision 0 Supplement 1 Analysis frequencies selected for each of the subgrade profile cases in the seismic SSI analyses are provided in Section 3C.5. Methodology used to calculate the Case 1 and Case 2 Seismic responses is summarized in Subsection 3C.1.1 through Subsection 3C.1.3.

3C.1.1 Stress Calculation Shell element forces and moments obtained from the Case 1 and Case 2 seismic SSI analyses are provided at the center of the element with respect to the element local coordinate system. To calculate the co-directional stress demands, the time histories of responses due to the three earthquake components are combined algebraically for each element using the time-step-by-time-step approach as discussed in Subsection 3.7.2.6. The combined stress demands are then transformed from the local coordinate system into a global cartesian or cylindrical coordinate system for use as input for the design of the structural members.

Average stress results from the corresponding five sets of Acceleration Time Histories (ATHs) represent the response of the structure for each of the SSI analysis cases considered. Stress results obtained from the analyses of the three subgrade profiles for Case 1 and Case 2, and from the sensitivity analysis discussed in Subsection 3C.4.3, are enveloped to provide enveloping CRN site-specific seismic stress demands that account for the uncertainties and variations of the subgrade properties and the seismological conditions at the site.

3C.1.2 In-Structure Response Spectra Calculation Per Subsection 3.7.2.5.1, the In-Structure Response Spectra (ISRS) are calculated at 301 frequency points equally distributed on the logarithmic scale at the frequency range from 0.1 Hz to 100 Hz to ensure the ISRS are sufficiently close to the peak response frequencies of the supporting structure. The ISRS are calculated at equipment locations for multiple damping ratios spanning the range corresponding to that of equipment under consideration.

Each SSI analysis case is performed separately for each one of the three directional components of input ground motion using five sets of ATHs. For each set, the ISRS obtained from the analysis of each of the three ground motion components are combined to get the total co-directional response using the Square-Root-of-the-Sum-of-Squares (SRSS) method. For each subgrade profile analysis case, average spectra from the corresponding sets of ATHs are computed representing the amplitude and frequency content of the in-structure motion at a particular location as illustrated in Figure 3C-1. The ISRS are calculated as the envelope of the results obtained from the Case 1 and Case 2 SSI analysis cases. To address uncertainties related to the modeling of natural frequencies of the supporting structure and the SSI analysis methodology, peaks of the enveloping ISRS are broadened by the 15% requirement specified in Subsection 3.7.2.5.1 as illustrated in Figure 3C-2. Following the guidelines of ASCE/SEI 4-16, Seismic Design of Safety-Related Nuclear Structures, (Reference 3C-3), sharp valleys between peaks may also be filled for use in the design and qualification of SSCs.

If the comparisons of the sensitivity analysis results discussed in Subsection 3C.4.3 and Section 3C.7 indicate that the seismic in-structure responses exceed the design basis seismic responses obtained from the Case 1 and Case 2 seismic SSI analyses by more than 10%, the results of the sensitivity analyses are also included in the design basis ISRS.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-3 Revision 0 Supplement 1 3C.1.3 Relative Displacement Calculation Relative displacements are calculated for key nodal locations throughout the RB relative to the free field at mat foundation elevation. Relative displacements are calculated using the same methodology used for ISRS calculation described in Subsection 3C.1.2, excluding broadening which is not applicable.

3C.2 Key Nodal Locations As stated in Subsections 3.7.2.2 and 3.7.2.9, responses at key nodal locations are used to demonstrate the accuracy of the seismic SSI analysis results, to determine analysis cutoff frequencies, and to evaluate the effects of different site conditions on the RB SSI response based on comparisons of results from the different seismic SSI analysis cases.

Key nodal locations selected per the criteria provided in Section 5.3.1 of NEDO-33914-A, BWRX-300 Advanced Civil Construction and Design Approach, (Reference 3C-4) for the verification and validation of the integrated RB seismic SSI analyses and results are provided in Table 3C-1.

3C.3 Key Structural Members Seismic stress demands are calculated for key structural members and compared between the different SSI analysis cases to understand the effects of site conditions on the seismic response of the soil-structure system.

Table 3C-2 presents the key structural members selected based on their critical support function of Safety Class 1 equipment, critical importance within the lateral force resisting system, or critical location along the seismic load path. These key structural members are also illustrated in Figure 3C-3.

3C.4 Seismic Analysis Cases The seismic analysis cases performed for the integrated RB are presented in Table 3C-3 and discussed in Subsection 3C.4.1 through Subsection 3C.4.3.

The envelope of the responses obtained from the six primary SSI analysis cases presented in Table 3C-3 forms the design basis for the BWRX-300 integrated RB at the CRN site. The other six analysis cases presented in Table 3C-3 are the sensitivity analysis cases performed to evaluate the effects of uncertainties and different site conditions discussed in Subsection 3.7.2.9.

3C.4.1 Case 1 The Case 1 seismic SSI analyses are performed using the standalone integrated RB FE model presented in Figure 3B-5 evaluated using the input response spectra presented in Figure 3.7-3 through Figure 3.7-5 and the three sets of subgrade profiles presented in Figure 3.7-25 and Figure 3.7-26.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-4 Revision 0 Supplement 1 The concrete backfill and mud mat dynamic properties assigned to the SSI FE model used in the Best Estimate (BE) Case 1, Lower Bound (LB) Case 1 and Upper Bound (UB) Case 1 analyses are presented in Table 3C-4.

3C.4.2 Case 2 The Case 2 seismic SSI analyses are performed using the combined SSI FE model presented in Figure 3B-7 evaluated using the Foundation Input Response Spectra (FIRS) at the bottom of the integrated RB foundation, the Performance Based Surface Response Spectra (PBSRS) at the plant finished grade elevation and the Performance Based Intermediate Response Spectra (PBIRS) at top of the un-weathered rock presented in Figure 3C-4.

Figure 3C-5 and Figure 3C-6 present the strain-compatible shear wave velocity profiles, compression wave velocity and damping ratio profiles used for the Case 2 seismic analyses of the integrated RB.

The concrete backfill and mud mat dynamic properties assigned to the SSI FE model used in the BE Case 2, LB Case 2 and UB Case 2 analyses are the same as those for Case 1 presented in Table 3C-4.

3C.4.3 Sensitivity Cases The six sensitivity analyses listed in Table 3C-3 and discussed in Subsection 3C.4.3.1 through Subsection 3C.4.3.6 are performed using the combined SSI FE model presented in Figure 3B-7 and the BE Case 2 subgrade profile.

3C.4.3.1 Cracked Sensitivity Analysis In this analysis, the BE Case 2 model is modified by considering that all concrete made members are fully cracked. Higher SSE-level damping values are assigned to the structural members reflecting higher energy dissipation associated with cracking of concrete. BE soil and rock properties and BE properties of the concrete fill are used. This case is performed to evaluate the variability in structural stiffness on the RB seismic response by considering LB structural stiffness for the RB.

3C.4.3.2 Excavation Support Sensitivity Analysis Given that the concrete fill around the RB is included in the BE Case 2 model, the uncertainty related to the excavation support provided by the temporary excavation support is investigated by evaluating the base case model supplemented by the explicit modeling of the temporary excavation support. Solid elements are used to represent the BE dynamic properties of the excavation support corresponding to a concrete compressive strength of 3,500 psi, which is the same as that of the BE concrete backfill properties.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-5 Revision 0 Supplement 1 3C.4.3.3 No-Friction Sensitivity Analysis The technique used for the RB shaft construction and corrosion allowance can affect the friction at the interfaces between the RB exterior wall and the surrounding concrete fill material. This case is performed to investigate the potential effects on the RB seismic response when, for short instances of time, parts of the below-grade concrete backfill may slip in tangential or vertical directions with respect to the RB exterior wall. The analysis is performed by modifying the BE Case 2 model such that it simulates a complete lack of traction between the concrete backfill and the exterior wall of the RB from ground surface to the bottom of the foundation.

3C.4.3.4 Dry Sensitivity Analysis The seismic design of the RB is based on the analysis of models that reflect fully saturated conditions for soil materials located below the nominal groundwater elevation considered to be located at finished grade as discussed in Subsections 3.7.1.3 and 3.7.2.9.3. The potential effects of groundwater level variability on the seismic design are addressed by comparing the seismic responses obtained from the SSI analysis of the dry soil profile. The dry soil profile uses BE soil dynamic properties representative of the conditions of the LB groundwater level at the site. In this analysis, the groundwater is conservatively considered to be located below the RB foundation bottom elevation, thus unsaturated soil Poissons ratios are used.

3C.4.3.5 Separation Sensitivity Analysis The separation sensitivity analysis is performed to investigate the effects on the RB seismic response when, for short instances of time, parts of the below-grade concrete backfill may separate from the surrounding subgrade soil. This separation would occur on one side of the backfill and along a portion of the embedment depth. The analysis is performed by modifying the BE Case 2 model such that it simulates a complete separation between the concrete backfill and the surrounding top layers of subgrade soil. The separation is conservatively considered for the full depth of soil, down to the top of un-weathered rock, and for the entire potential separation surface around the backfill.

3C.4.3.6 Lean Concrete Sensitivity Analysis The use of concrete fill as the design basis analysis case introduces the additional uncertainty over the strength of the concrete fill which is not designed to maintain its structural integrity throughout the entire operational life of the plant. To address this uncertainty, a sensitivity evaluation is performed using a reduced value of the design concrete fill strength corresponding to a concrete compressive strength of 1,500 psi.

3C.5 Design Basis Seismic Analysis Cases and Selected Frequencies of Analysis As stated in Section 3C.4, the six primary SSI analysis cases presented in Table 3C-3 form the design basis for the integrated RB at the CRN Site.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-6 Revision 0 Supplement 1 The six primary seismic SSI analysis cases are performed for frequencies starting at 0.012 Hz to the respective cutoff frequency for each of the subgrade profiles. Following the selection criteria in Section 5.3.2 of NEDO-33914-A and as indicated in Table 3C-5, the following cutoff frequencies are adopted for the analysis:

35 Hz for the LB Cases 1 and 2

51 Hz for the BE Cases 1 and 2

70 Hz for the UB Cases 1 and 2 For the UB Cases 1 and 2, the 70 Hz cutoff frequency allows for the adequate calculation of ISRS up to a minimum of 50 Hz.

The SASSI frequency domain analysis calculates the response of the system at selected Structure frequencies defined as the number of frequencies necessary to adequately characterize the response of the full SSI system. The results for the response at frequencies between the selected frequencies are interpolated.

The calculated frequencies are narrowly spaced with a total of 197 frequencies for the UB case, which ensures that Acceleration Transfer Function (ATF) peaks and troughs are adequately captured and minimizes ATF interpolation concerns.

3C.5.1 Subgrade Properties Variation Effects Figure 3C-7 compares the ATFs representing the co-directional responses at the RB exterior wall at grade in the three orthogonal directions obtained from the six primary SSI analysis cases.

The plots in Figure 3C-7 show distinct dominant frequencies for the SSI system depending on the subgrade profile analysis. Following the expected trend, the ATF peak frequencies obtained from the LB profile are associated with the lowest frequency, followed by higher ATF peak frequencies for the BE profile and with the highest ATF peak frequency associated with the UB profile as shown in Figure 3C-7. The dots in the ATF plots in Figure 3C-7 represent the calculated Structure frequencies, and the smooth curves in between show the interpolated ATFs.

Figure 3C-8 compares the ATF results obtained from the analysis of the six primary and six sensitivity cases listed in Table 3C-3. The peak at the smallest frequency of around 3.5 Hz is associated with the No-Friction sensitivity case. Figure 3C-9 presents the ATF normalized by the ATF of the free-field outcrop motion corresponding to the RB exterior wall at grade. These ATFs are normalized with respect to the free-field response at surface therefore removing the large amplitude ATF peaks at the embedment soil column frequencies to better depict the peak responses of the RB structure. The amplitudes of the normalized ATFs fall well below unity at frequencies that are smaller than their respective cut-off frequencies thus providing confirmation of the adequacy of the selected cut-off frequencies of analysis. The exception in the LB Case 1 normalized ATF is due to the amplitude of the free-field ATF (used in the denominator to calculate the normalized ATF) approaching zero at frequencies larger than 20 Hz, which results in the artificially high amplitude of the normalized ATF.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-7 Revision 0 Supplement 1 3C.6 Seismic Response Characteristics of the Integrated Reactor Building at the Clinch River Nuclear Site 3C.6.1 Seismic Structural Stress Responses for Cases 1 and 2 This section presents plots of main stress components in key structural members to depict the characteristics of how the integrated RB structures resist the CRN site-specific seismic loads.

Member force and moment results obtained from the analysis of the primary SSI cases are compared to evaluate the effect of geotechnical and seismological site conditions on the seismic stress responses of the integrated RB structures.

Figure 3C-10 through Figure 3C-21 compare the in-plane membrane forces and out-of-plane bending moments on the below grade portions of the RB exterior wall obtained from the LB, BE and UB Cases 1 and 2. Due to the cylindrical shape of the wall, the out-of-plane dynamic pressure forces applied on the below grade portion of the RB exterior wall are resisted by the tangential (hoop) membrane forces as shown in Figure 3C-10 through Figure 3C-12 and in-plane shear forces as shown in Figure 3C-13 through Figure 3C-15. Comparing the hoop forces due to the LB, BE, and UB cases in Figure 3C-10 through Figure 3C-12, the LB case is found to provide the largest seismic demands followed by the BE and UB cases. This can be attributed to the larger dynamic earth pressures applied on the deeply embedded structure supported by the softer soil layers. A similar pattern is observed for the in-plane shear forces shown in Figure 3C-13 through Figure 3C-15, and vertical forces shown in Figure 3C-16 through Figure 3C-18. The plots in Figure 3C-19 through Figure 3C-21 show that the out-of-plane horizontal moments are of a small magnitude with peak amplitudes at the connections of the RB exterior wall with the interior floors and wing walls.

The RB cylindrical shape helps reduce the demands on the RB wing walls that run in the radial direction connecting the RB exterior wall to the SCCV wall. This is demonstrated by the in-plane shear forces in wing walls illustrated in Figure 3C-22 through Figure 3C-24.

3C.6.2 Seismic In-Structure Responses for Cases 1 and 2 Four percent (4%) damped ISRS for responses at key nodal locations obtained from the analysis of the six primary SSI cases and the six sensitivity cases (see Section 3C.7) are compared as shown in Figure 3C-25 for the integrated RB common mat foundation to evaluate the effect of subgrade variations on the RB in-structure response.

At the mat foundation, the response is dependent on the subgrade profile, as can be seen by the increasing peak frequencies with increasing profile stiffness in Figure 3C-25. The BE Case 2 and the six sensitivity cases which share the same subgrade properties have similar in-structure responses at the center of the common mat foundation, with the exception of the vertical ISRS for the Dry case. This exception is due to the differences in the compression wave velocities of the top 37.4 ft of soil layers.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-8 Revision 0 Supplement 1 3C.7 Sensitivity Analysis Cases As stated in Subsection 3.7.2.9, sensitivity evaluations are performed to investigate the effects of structural stiffness variation, soil-structure interface conditions, soil separation, groundwater variations, and excavation and backfill properties on the seismic response and design of the CRN RB SSCs.

The sensitivity evaluations are based on results of sensitivity SSI analyses performed using the BE subgrade profile properties and models of the combined SSI model shown in Figure 3B-7 representing conditions that bound the variation of the evaluated SSI parameter. The sensitivity evaluations are performed by comparing in-structure responses at key locations and stress demands for key structural members within the RB. Subsection 3C.7.1 through Subsection 3C.7.4 summarize the findings of these sensitivity evaluations.

Per guidance of Section 5.3 of NEDO-33914-A, the results of the sensitivity analysis are included in the RB seismic design basis when comparisons show significant exceedances, greater than 10%, in the RB seismic response due to any of the examined effects. To do so, results of the sensitivity cases are enveloped with those of the Case 2 design basis analyses of LB, BE, and UB subgrade profiles. The enveloped results are then used to develop seismic member force and moment demands, and design ISRS, following the methodology in Subsection 3C.1.1 and Subsection 3C.1.2 for the Seismic Category I SSCs at the CRN Site.

In general, the comparisons show that seismic forces and moments do not differ significantly from the base case BE Case 2 with some exceptions, notably for the No-Friction and Cracked sensitivity cases as demonstrated in Subsection 3C.7.1 through Subsection 3C.7.5.

3C.7.1 Structural Stiffness and Damping Variation Effects on Responses Figure 3C-26 through Figure 3C-30 compare stress responses obtained from the cracked sensitivity analysis discussed in Subsection 3C.4.3.1 and the BE Case 2 analysis discussed in Subsection 3C.4.2 at the RB exterior wall, SCCV top slab and RPV pedestal.

The comparisons of stress responses obtained from the models with BE (Effective) and LB (Reduced) stiffness properties at the RB exterior wall, the RPV pedestal and at the SCCV top slab show a noticeable decrease in stress demands due to the reduction in stiffness. In the case of the RB exterior wall, the decrease in stress demands is around 35% for vertical forces (Figure 3C-26) and 25% for in-plane shear forces (Figure 3C-27). The out-of-plane moment decrease in the SCCV top slab ranges from 10% to 15% (Figure 3C-28), while the RPV pedestal demand decrease is approximately 20% for vertical (Figure 3C-29) and 15% for in-plane shear forces (Figure 3C-30).

3C.7.2 Excavation Support, Backfill and Friction Effects on Responses Figure 3C-31 through Figure 3C-35 compare stress responses obtained from the excavation support sensitivity analysis discussed in Subsection 3C.4.3.2 and the BE Case 2 analysis discussed in Subsection 3C.4.2 at the RB exterior wall, SCCV top slab and RPV pedestal. The comparisons show that the overall effects of excavation support on these structures are small.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-9 Revision 0 Supplement 1 Figure 3C-36 through Figure 3C-40 compare stress responses obtained from the no-friction sensitivity analysis discussed in Subsection 3C.4.3.3 and the BE Case 2 analysis discussed in Subsection 3C.4.2 at the RB exterior wall, SCCV top slab and RPV pedestal. The comparisons show a large increase in vertical forces in the below-grade segment of the RB exterior wall (about 4 times that of BE Case 2, see Figure 3C-36) due to the absence of frictional bond between the wall and the surrounding concrete backfill. An increase of about 30% (Figure 3C-38) is also observed in the SCCV top slab out-of-plane moments. Similarly, for the RPV pedestal, Figure 3C-39 and Figure 3C-40 indicate an increase of about 25% for vertical and 15% for in-plane forces.

3C.7.3 Groundwater Variation Effects on Responses Figure 3C-41 through Figure 3C-45 compare stress responses obtained from the Saturated and dry subgrade profiles (see Subsection 3C.4.3.4) at the RB exterior wall, SCCV top slab and RPV pedestal.

The comparisons show that the overall effect of groundwater variation on the structures is small, with the largest effect being an increase in the SCCV top slab out-of-plane moments of about 15% (Figure 3C-43).

3C.7.4 Soil Separation Effects on Responses Figure 3C-46 through Figure 3C-50 compare stress responses obtained from the separation sensitivity analysis discussed in Subsection 3C.4.3.5 and the BE Case 2 analysis discussed in Subsection 3C.4.2 at the RB exterior wall, SCCV top slab and RPV pedestal.

The comparisons show that the overall effect of soil separation on the structures is small with the largest increase being in the order of 10% in vertical forces in the below-grade segment of the RB exterior wall (Figure 3C-46).

3C.7.5 Concrete Fill Strength Effects on Responses Figure 3C-51 through Figure 3C-55 compare stress responses obtained from the lean concrete sensitivity analysis discussed in Subsection 3C.4.3.6 and the BE Case 2 analysis discussed in Subsection 3C.4.2 at the RB exterior wall, SCCV top slab and RPV pedestal.

The comparisons show a 10% to 15% increase in the RB wall vertical and in-plane shear forces due to the reduced strength of the concrete backfill (Figure 3C-51 and Figure 3C-52) but no increase on stress responses for the SCCV top slab or RPV pedestal (Figure 3C-53 through Figure 3C-55).

3C.8 SSE Structural Stress Demands To illustrate the seismic demands on the Seismic Category I integrated RB structures, critical seismic member force and moment demands are presented in this section for key structural members.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-10 Revision 0 Supplement 1 3C.8.1 Containment Stress Demands Figure 3C-56 presents the seismic moment demands for the SCCV top slab, while Figure 3C-57 and Figure 3C-58 present the bounding seismic membrane forces and moment demands for the SCCV wall.

3C.8.2 Containment Internal Structures Stress Demands Figure 3C-59 presents the seismic membrane forces for the RPV pedestal.

3C.8.3 Reactor Building Stress Demands Figure 3C-60 and Figure 3C-61 present the bounding seismic membrane forces and moment demands for the RB exterior wall.

3C.9 Maximum SSE Displacements Table 3C-6 presents the amplitudes of seismic displacements at key locations within the integrated RB structures. These displacements are computed relative to the free-field at the bottom of the RB foundation and are an envelope of results obtained from the six primary SSI analysis cases and six sensitivity cases described in Section 3C.4.

3C.10 References 3C-1 ASCE/SEI 43-19, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, American Society of Civil Engineers, 2019.

3C-2 RG 1.61, Damping Values for Seismic Design of Nuclear Power Plants, U.S. Nuclear Regulatory Commission.

3C-3 ASCE/SEI 4-16, Seismic Design of Safety-Related Nuclear Structures, American Society of Civil Engineers, 2017.

3C-4 NEDO-33914-A, BWRX-300 Advanced Civil Construction and Design Approach, GE-Hitachi Nuclear Energy Americas, LLC, Revision 2, June 2022.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-11 Revision 0 Supplement 1 Table 3C-1 Key Nodal Locations Node Number Location Description Elevation

[m (ft)]

54526 Center of RB mat foundation

-35.22 (-115.6) 57403 Edge of RB mat foundation

-35.22 (-115.6) 97673 RPV support at top of pedestal

-8.95 (-29.4) 88881 RB exterior wall at El. -15m (-49 ft) floor

-14.805 (-48.6) 122772 RB exterior wall at grade 0.00 (0.00) 141410 RB exterior wall at operating floor 14.238 (46.7) 146384 Edge of RB roof 29.115 (95.5) 150753 Center of RB roof 32.258 (105.8)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-12 Revision 0 Supplement 1 Table 3C-2 Key Structural Members Structural Member Description Critical Stress Components Mat Foundation Slab Circular slab Out-of-plane moment Steel-Plate Composite Containment Vessel (SCCV)

Top slab Circular slab Out-of-plane moment SCCV wall Cylindrical wall Membrane forces, in-plane shear, out-of-plane moment RPV pedestal Cylindrical wall Membrane forces, in-plane shear RB exterior wall Cylindrical wall Membrane forces, in-plane shear, out-of-plane moment Wing walls Rectangular walls Membrane forces, in-plane shear

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-13 Revision 0 Supplement 1 Table 3C-3 Seismic Analysis Cases Analysis Case Soil Profile Category BE Case 1 BE Case 1 Primary LB Case 1 LB Case 1 Primary UB Case 1 UB Case 1 Primary BE Case 2 BE Case 2 Primary LB Case 2 LB Case 2 Primary UB Case 2 UB Case 2 Primary Separation BE Case 2 Sensitivity No-Friction BE Case 2 Sensitivity Excavation Support BE Case 2 Sensitivity Lean Concrete BE Case 2 Sensitivity Dry BE Case 2-Dry Sensitivity Cracked BE Case 2 Sensitivity

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-14 Revision 0 Supplement 1 Table 3C-4 Concrete Fill Dynamic Properties Case Shear-wave Velocity

[ft/s]

Compression-wave Velocity

[ft/s]

Damping Ratio [%]

BE 6673 10580 1.0 LB 6135 9731 1.8 UB 7106 11270 0.5

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-15 Revision 0 Supplement 1 Table 3C-5 Dominant Frequencies for Soil-Structure InteractionSystem Response Soil Case Vs at FIRS [ft/sec]

Cut-off Frequency

[Hz]

Case 1 Case 2 LB Cases 1 and 2 6,122 6,194 35.16 BE Cases 1 and 2 7,549 7,621 51.27 UB Cases 1 and 2 9,311 9,380 70.19 BE Case 2-Dry N/A 7,621 51.27

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-16 Revision 0 Supplement 1 Table 3C-6 Enveloped Seismic Relative Displacements at Key Locations Node Number Location Description Seismic Relative Displacement (in)

X Y

Z 54526 Center of RB mat foundation 0.01 0.01 0.01 57403 Edge of RB mat foundation 0.01 0.01 0.01 97673 RPV support at top of pedestal 0.22 0.23 0.04 88881 RB exterior wall at El. -15m (-49ft) floor 0.19 0.17 0.13 122772 RB exterior wall at grade 0.20 0.27 0.22 141410 RB exterior wall at operating floor 0.46 0.55 0.27 146384 Edge of RB roof 0.72 0.86 0.28 150753 Center of RB roof 0.76 0.95 0.55

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-17 Revision 0 Supplement 1 Figure 3C-1 Averaging of In-Structure Response Spectra from Five Sets of Acceleration Time History Results

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-18 Revision 0 Supplement 1 Figure 3C-2 Enveloping and Broadening of In-Structure Response Spectra

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-19 Revision 0 Supplement 1 Figure 3C-3 Reactor Building Finite Element Model

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-20 Revision 0 Supplement 1 Figure 3C-4 Case 2 Horizontal and Vertical Input Response Spectra at Bottom of Reactor Building Foundation, at Top of Un-Weathered Rock and at Plant Finished Grade Elevation

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-21 Revision 0 Supplement 1 Figure 3C-5 Case 2 Hazard Consistent Strain-Compatible Shear Wave Velocity (Vs) and Compression (P)-Wave Velocity (Vp) Profiles

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-22 Revision 0 Supplement 1 Figure 3C-6 Case 2 Hazard Consistent Strain-Compatible Damping Ratios

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-23 Revision 0 Supplement 1 Figure 3C-7 Effects of Subgrade Variation on the Acceleration Time Histories - Reactor Building Exterior Wall at Grade

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-24 Revision 0 Supplement 1 Figure 3C-8 Comparison of Acceleration Transfer Function for Primary and Sensitivity Cases - Reactor Building Exterior Wall at Grade

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-25 Revision 0 Supplement 1 Figure 3C-9 Effects of Subgrade Variation on Normalized Acceleration Transfer Function -

Reactor Building Exterior Wall at Grade

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-26 Revision 0 Supplement 1 Figure 3C-10 Membrane Tangential (Hoop) Forces - Exterior Wall Below Grade - Lower Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-27 Revision 0 Supplement 1 Figure 3C-11 Membrane Tangential (Hoop) Forces - Exterior Wall Below Grade - Best Estimate Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-28 Revision 0 Supplement 1 Figure 3C-12 Membrane Tangential (Hoop) Forces - Exterior Wall Below Grade - Upper Bound Case 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-29 Revision 0 Supplement 1 Figure 3C-13 In-plane Shear Forces - Exterior Wall Below Grade - Lower Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-30 Revision 0 Supplement 1 Figure 3C-14 In-plane Shear Forces - Exterior Wall Below Grade - Best Estimate Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-31 Revision 0 Supplement 1 Figure 3C-15 In-plane Shear Forces - Exterior Wall Below Grade - Upper Bound Case 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-32 Revision 0 Supplement 1 Figure 3C-16 Membrane Vertical Forces - Exterior Wall Below Grade - Lower Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-33 Revision 0 Supplement 1 Figure 3C-17 Membrane Vertical Forces - Exterior Wall Below Grade - Best Estimate Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-34 Revision 0 Supplement 1 Figure 3C-18 Membrane Vertical Forces - Exterior Wall Below Grade - Upper Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-35 Revision 0 Supplement 1 Figure 3C-19 Out-of-Plane Horizontal Moments - Exterior Wall Below Grade - Lower Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-36 Revision 0 Supplement 1 Figure 3C-20 Out-of-Plane Horizontal Moments - Exterior Wall Below Grade - Best Estimate Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-37 Revision 0 Supplement 1 Figure 3C-21 Out-of-Plane Horizontal Moments - Exterior Wall Below Grade - Upper Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-38 Revision 0 Supplement 1 Figure 3C-22 In-Plane Shear Forces - Wing Walls - Lower Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-39 Revision 0 Supplement 1 Figure 3C-23 In-Plane Shear Forces - Wing Walls - Best Estimates Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-40 Revision 0 Supplement 1 Figure 3C-24 In-Plane Shear Forces - Wing Walls - Upper Bound Cases 1 and 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-41 Revision 0 Supplement 1 Figure 3C-25 In-Structure Response at Center of Common Mat Foundation

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-42 Revision 0 Supplement 1 Figure 3C-26 Structural Stiffness Variation Effects on Vertical Forces in Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-43 Revision 0 Supplement 1 Figure 3C-27 Structural Stiffness Variation Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-44 Revision 0 Supplement 1 Figure 3C-28 Structural Stiffness Variation Effects on Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-45 Revision 0 Supplement 1 Figure 3C-29 Structural Stiffness Variation Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-46 Revision 0 Supplement 1 Figure 3C-30 Structural Stiffness Variation Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-47 Revision 0 Supplement 1 Figure 3C-31 Excavation Support Effects on Vertical Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-48 Revision 0 Supplement 1 Figure 3C-32 Excavation Support Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-49 Revision 0 Supplement 1 Figure 3C-33 Excavation Support Effects on Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-50 Revision 0 Supplement 1 Figure 3C-34 Excavation Support Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-51 Revision 0 Supplement 1 Figure 3C-35 Excavation Support Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-52 Revision 0 Supplement 1 Figure 3C-36 Friction Effects on Vertical Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-53 Revision 0 Supplement 1 Figure 3C-37 Friction Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-54 Revision 0 Supplement 1 Figure 3C-38 Friction Effects on the Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-55 Revision 0 Supplement 1 Figure 3C-39 Friction Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-56 Revision 0 Supplement 1 Figure 3C-40 Friction Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-57 Revision 0 Supplement 1 Figure 3C-41 Groundwater Variation Effects on Vertical Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-58 Revision 0 Supplement 1 Figure 3C-42 Groundwater Variation Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-59 Revision 0 Supplement 1 Figure 3C-43 Groundwater Variation Effects on Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-60 Revision 0 Supplement 1 Figure 3C-44 Groundwater Variation Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-61 Revision 0 Supplement 1 Figure 3C-45 Groundwater Variation Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-62 Revision 0 Supplement 1 Figure 3C-46 Separation Effects on Vertical Forces in the Reactor Building External Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-63 Revision 0 Supplement 1 Figure 3C-47 Separation Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-64 Revision 0 Supplement 1 Figure 3C-48 Separation Effects on Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-65 Revision 0 Supplement 1 Figure 3C-49 Separation Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-66 Revision 0 Supplement 1 Figure 3C-50 Separation Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-67 Revision 0 Supplement 1 Figure 3C-51 Concrete Fill Strength Effects on Vertical Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-68 Revision 0 Supplement 1 Figure 3C-52 Concrete Fill Strength Effects on In-Plane Shear Forces in the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-69 Revision 0 Supplement 1 Figure 3C-53 Concrete Fill Strength Effects on Containment Top Slab Out-of-Plane Moment

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-70 Revision 0 Supplement 1 Figure 3C-54 Concrete Fill Strength Effects on Vertical Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-71 Revision 0 Supplement 1 Figure 3C-55 Concrete Fill Strength Effects on In-Plane Shear Forces in the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-72 Revision 0 Supplement 1 Figure 3C-56 Seismic Design Basis Moment Demands for the Containment Top Slab

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-73 Revision 0 Supplement 1 Figure 3C-57 Seismic Design Basis Membrane Forces for the Containment Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-74 Revision 0 Supplement 1 Figure 3C-58 Seismic Design Basis Moments for the Containment Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-75 Revision 0 Supplement 1 Figure 3C-59 Seismic Design Basis Membrane Forces for the Reactor Pressure Vessel Pedestal

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-76 Revision 0 Supplement 1 Figure 3C-60 Seismic Design Basis Membrane Forces for the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3C-77 Revision 0 Supplement 1 Figure 3C-61 Seismic Design Basis Moments for the Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3D-1 Revision 0 Supplement 1 APPENDIX 3D INTERACTION EVALUATIONS RESULTS 3D.1 Introduction This appendix provides interaction evaluations results of the Seismic Category II and RW-IIa Power Block structures and foundations with the Seismic Category I structures and foundations under extreme loading.

Appendix 3D will be supplied in a Supplement in the 4th quarter of FY 2025.Implementation of the methodology for these evaluations discussed in Section 3.3 and Section 3.7 requires the design of the surrounding Seismic Category II and RW-IIa Power Block structures and foundations to be completed to provide quantifiable results, and, thus, the interaction evaluations will be provided to support the FSAR.

Supplement 1, LCR 25-050

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-1 Revision 0 Supplement 1 APPENDIX 3E RESPONSES TO STATIC LOADS 3E.1 Introduction This appendix presents key design parameters and key results of the 1 g static SSI, static and quasi static, and thermal load cases considered bounding for the design of the BWRX 300 integrated Reactor Building structures to identify Critical demands on the structures.

Appendix 3E will be supplied in a Supplement in the 4th quarter of FY 2025.

Supplement 1, LCR 25-050 This appendix documents the 1-g static SSI, static and quasi-static and thermal analyses performed for the design of the integrated Reactor Building (RB) structures. The analyses are performed per the methodology presented in Subsection 3.8.4.1.4, using the Finite Element (FE) models described Section 3B.3 through Section 3B.5, and design parameters discussed in Section 3E.1.

The 1-g static SSI, static and quasi-static and thermal analyses are performed for the applicable loads listed in Section 3.8, excluding construction, refueling and piping reaction loading. These loads are judged to be non-governing for the design of the integrated RB structures and will be considered in the final design of the structures. Tornado missiles loadings are also not considered in the analyses as the thickness of the RB exterior wall above grade and the roof thickness is 2-3 times larger than the minimum acceptable barrier thickness requirements listed in Table 3.5-2. Tornado missile loading effects are also not expected to govern the design of the integrated RB and this assumption will be validated in the final design of structures.

Beyond design basis events are also not included in the load cases summarized in this appendix.

3E.1 Design Parameters Considered in the Structural Analyses The following are the design parameters considered in the 1-g static SSI, static and quasi-static and thermal analyses. For live loads considered in the analyses, refer to Subsection 3B.1.7.2.

3E.1.1 Equivalent Linear Static Subgrade Properties The site-specific geotechnical parameters presented in Table 3.8-8 and Table 3.8-9 are used to develop the input for the 1-g static SSI and subgrade impedance analyses used to evaluate the structural response to significant static and thermal design loads.

Table 3E-1 identifies the equivalent linear static subgrade profiles used for each of the structural analyses. The Lower Bound (LB) and Upper Bound (UB) elastic moduli (

) and Poissons ratios (

) in Table 3E-1 are the LB and UB as-built static subgrade properties provided in Table 3.8-8 and Table 3.8-9. The effective soil unit weight ( ) is calculated as the total soil/rock unit weights provided in Table 3.8-8 and Table 3.8-9 minus the unit weight of water.

Est st

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-2 Revision 0 Supplement 1 LB subgrade properties are used for the static and quasi-static analyses to emphasize the deformations at the subgrade-structure interfaces resulting in UB estimates of member forces and moment demands. Both LB and UB subgrade stiffness properties and are considered for the thermal stress analyses to bound the thermal stress demands on the RB integrated structures.

To evaluate the RB exterior wall, the 1-g SSI analyses use UB soil unit weight, LB and UB to emphasize the lateral earth pressures applied from the subgrade on the RB exterior wall.

3E.1.2 Nominal Groundwater Table Elevation The groundwater at the Clinch River Nuclear (CRN) site is located at plant finished grade (Elevation 814.5 ft) as per Table 2.0-1R.

3E.1.3 Additional Horizontal Rock Pressure Residual horizontal stresses that exceed the magnitude of the vertical stresses may be present in the rock masses at the CRN site due to past seismic activities. The results of past studies indicate these stresses are anisotropic, i.e., the magnitude of the pressures in the two orthogonal horizontal directions is different.

Rock excavation can relief the residual horizontal stresses due to deformation of the rock mass.

However, the magnitude of the rock mass displacements and resulting relief of residual rock stresses depends on the duration of the excavation and rock reinforcement that may be used to secure the excavation. The durability of the temporary rock reinforcement elements during the life of the plant is not guaranteed. If the rock reinforcement decays with time, the in-situ residual stress not already relieved during excavation can transfer from the reinforcement to the RB exterior wall. As a result, residual rock pressures, ranging circumferentially between 43.5 psi and 14.5 psi, are applied on the portion of the RB embedded in rock to account for the additional rock pressure from unstable rock. These pressures are applied uniformly along the depth of the embedded RB exterior wall starting at depth of 12.9 m (42.3 ft) below plant grade down to the bottom of the mat foundation but non-symmetrically around the RB circumference within the rock layers.

3E.1.4 Ambient Temperature The stress-free temperature for the design of the RB and Steel-Plate Composite Containment Vessel (SCCV) at the CRN site is 60°F.

The design temperatures representing the extreme external atmospheric conditions considered in the thermal analyses are -40°F in winter and 104°F in summer. Although the governing summer ambient temperature at the CRN Site is 107°F per Table 2.0-1R, the 3°F difference is not significant as thermal stress demands from the summer condition do not govern the thermal stress demands from winter conditions which create the most severe thermal gradients on steel-plate composite components.

Est st Est st

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-3 Revision 0 Supplement 1 3E.1.5 Precipitation Design Parameters The CRN RB design considers a normal and extreme ground snow loading of 52.2 psf and 104 psf, respectively. These loads bound the CRN design-basis ground snow loads of 21.93 psf.

3E.1.6 Wind and Extreme Wind Parameters As stated in Subsection 3.3.1.1, the integrated RB is designed for a wind load of 160 mph based on the ASCE/SEI 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, (Reference 3E-1) Risk Category IV, basic wind speed maps (3-second gust). This wind speed bounds the design basis wind speeds including hurricane for the CRN site (130 mph) listed in Table 2.0-1R.

The design basis tornado parameters are those listed in Table 2.0-1R. Extreme wind-generated missiles considered in the design of the integrated RB are those listed in Table 3.5-1.

Tornado wind load calculations were performed using both the BC-TOP-3-A, Tornado and Extreme Wind Design Criteria for Nuclear Power Plants, (Reference 3E-2) approach as discussed in Section 3.3.2.2 and the ASCE/SEI 7-22, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, (Reference 3E-3) approach, with the conclusion that the latter provides bounding pressure parameters. The bounding parameters were applied to the analysis model.

3E.1.7 Pool Water Levels Pool water levels used for the hydrostatic load case are based on normal operating levels and are as follows:

Isolation condenser pool water depth is 23.6 ft

Equipment and reactor cavity pools water depth is 25.6 ft

Fuel pool water depth is 44.9 ft 3E.1.8 Crane Loading Table 3E-2 summarizes the crane loads applied on the crane supporting structures as nodal forces in the FE model.

3E.1.9 Containment Internal Pressure Loading The containment pressure loadings considered in the analyses are the normal operating pressure loads Po, the design basis accident pressure Pa1 generated by a Loss-of-Coolant Accident (LOCA) inside containment, and the test pressure load (Pt) presented in Table 3E-3.

Conservatively, the negative containment pressure (Po-) which does not occur under normal operation but can occur during Anticipated Operational Occurrences (AOOs) is considered for the normal operating load case as indicated in Table 3E-3.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-4 Revision 0 Supplement 1 Test pressure load (Pt) applied to the SCCV is defined as 1.15 times the containment accident pressure (Pa1).

3E.1.10 Internal Pressure Loading Outside Containment An AOO exists in which the isolation condenser pools will develop internal pressure as the pool water boils and the enclosed pools develop internal pressure. In the absence of more information about the event precipitating the AOO, an internal pressure in the isolation condenser pools (Pop, see Table 3E-3) of 4.5 psi is also considered in the analyses.

An accident inside the steam tunnel causing internal pressure Pa2 is also considered in the analyses as indicated in Table 3E-3.

3E.1.11 Internal Flooding The RB subgrade floors are not expected to flood simultaneously due to a water line rupture.

Based on preliminary design evaluations, flooding of the floor slab above the common mat foundation is observed to be the most critical. As a result, a lone flood depth of 4.9 ft is considered to act on the floor slab above the common mat foundation.

Figure 3E-1 presents the hydrostatic loads considered in the design of the integrated RB structures as a result of a containment flooding generated by LOCA.

3E.1.12 Thermal Conditions Thermal stress analyses are performed by applying temperature loads based on heat transfer analysis.

The thermal stress analyses consider two cases, winter and summer and four accident conditions, with two accident durations each, as indicated in Table 3B-1.

Table 3E-4 present the summer and winter thermal conditions inside and outside the integrated RB during normal operations.

Summer/winter thermal conditions inside containment during normal operations are taken as 149°F and for accident conditions, the maximum design internal temperature considered is 330°F.

3E.2 1-g Static Soil-Structure Interaction Analysis The 1-g static SSI analysis is performed in ANSYS and considers the interaction of the deeply embedded RB with the surrounding Power Block structures and subgrade under 1-g loads (See Appendix 3I).

The model used for the 1-g static SSI analyses is described Section 3B.3. The soil and rock layers of the model are assigned the equivalent linear elastic subgrade properties summarized in Table 3E-1. Structural elements of the model are assigned BE stiffness properties as discussed Section 3B.3. Submerged unit weight properties are assigned to the subgrade materials located

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-5 Revision 0 Supplement 1 below the bounding nominal groundwater level specified at plant finished grade as noted in Subsection 3E.1.2. Boundary conditions considered for the 1-g static SSI analysis are described in Table 3B-1.

3E.2.1 Analysis Cases Two sets of vertical 1-g static SSI analyses are performed for the unfactored and factored gravity inertia load conditions as discussed in Section 3B.3 and shown in Table 3B-1. Demands obtained from the 1-g static SSI analysis are from:

The dead load from the self-weight and weight of permanent equipment and components.

The vertical fluid load on pool slabs from the weight of the water in the pools.

The static earth pressure load due to submerged weight of the soil and rock.

The surcharge lateral pressure load from the surrounding Power Block surface mounted foundations conservatively captured by assigning higher mass density properties as discussed Section 3B.3.

Crane girders and trolley weights applied as vertical forces with the trolley parked at the end of the crane.

In the model used to analyze the factored load combination, the submerged unit weight properties of the subgrade and surrounding structures are increased by 33% to account for the higher factored lateral earth pressure loads also as described in Section 3B.3.

3E.2.2 Response and Demands from Dead Load and Earth Pressure Loads Figure 3E-2 and Figure 3E-3 present the magnitude of internal forces obtained from the 1-g SSI analysis in the RB exterior wall and SCCV wall, respectively.

3E.3 Static and Quasi-Static Analysis The analyses of static and quasi-static loads are performed on the FE models described Section 3B.4. As stated in Subsection 3.8.4.1.4, the analysis of the static and quasi-static loads that have global effect on the response of the integrated RB structural model are performed using the subgrade impedance sub-structuring method that accounts for the subgrade stiffness at the interfaces with the RB structure. Linear elastic contact springs connect the integrated RB structural model and subgrade FE models developed using the subgrade stiffness impedance methodology. Stiffness properties are assigned to the contact springs to adequately represent the interaction mechanism between the RB structure, backfill, and subgrade.

Analyses performed to determine force/moment demands due to load cases for which the interaction with the surrounding soil is not important are performed on models with prescribed boundary conditions.

Table 3B-1 summarizes the modeling requirements for the various static and quasi-static analysis cases required for the design of the integrated RB structures.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-6 Revision 0 Supplement 1 3E.3.1 Analysis Cases The static and quasi-static analysis cases performed are those described in Table 3B-1, excluding analysis cases associated with normal operating and Design Basis Accident (DBA) nozzle, equipment and piping reaction loads and local loads on containment during DBA for which information is not available.

The following subsections present responses and demands obtained from load cases governing the design of the integrated RB structures.

3E.3.2 Response and Demands from Horizontal Hydrostatic and Hydrodynamic Loads Horizontal pool water loads applied on the integrated RB model consist of the horizontal hydrostatic loads (Fh) and sloshing and breathing mode hydrodynamic pressures due to seismic (Safe Shutdown Earthquake (SSE)) excitation.

Hydrostatic pressures based on the pool water levels listed in Subsection 3E.1.7 are applied as an outward normal pressure with a gradient equal to the density of water at approximately 60°F (62.4 lb/ft3) multiplied by the distance from top of water. No pressure is applied on the intermediate pool walls that restrain fluid on both sides as the net force on the intermediate wall is negligible.

The deformed shape response of the pool walls due to hydrostatic loads is shown in Figure 3E-4(a). The outward expansion of the pool walls shown in Figure 3E-4(a) is consistent with the loading.

Hydrodynamic pressures on pool walls due to vertical earthquake excitation, referred to as breathing water pressure loads, are also applied as outward normal pressures on pool walls using the same approach as for hydrostatic pressures. The magnitude of the fluid pressure is scaled based on the vertical Zero Period Acceleration (ZPA) the pools experience presented in Table 3E-5. Breathing water pressures are applied on intermediate pool walls that restrain fluid from both sides as these walls may experience a net pressure depending on the differential ZPA in various regions.

As illustrated in Figure 3E-4(b), the outward expansion of the pool walls under breathing water pressures is also consistent with the loading.

Sloshing pressure is obtained based on the 0.5% damped spectral acceleration corresponding to the sloshing mode frequency at the pool elevation. Sloshing pressure is based on a bounding spectral acceleration of 0.5 g. Sloshing pressure loads are applied to the ANSYS model as nodal forces in the corresponding sloshing directions for X (North-South) and Y (East-West) loading directions. The sloshing pressure force magnitudes are dependent on the pool plan geometry, water depth, and seismic input. For the calculation of the sloshing pressures, the RB pools are divided in multiple regions of equivalent rectangular dimensions. For each pool region, the general sloshing parameters are calculated, then the pressure for each wall or floor shell is calculated by considering the relative water depth and location. Corner forces are calculated for each shell using the projected shell area in the sloshing direction.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-7 Revision 0 Supplement 1 3E.3.3 Responses and Demands from Containment Internal Pressure Load Cases The containment pressure loadings listed in Table 3E-3 are applied normal to the SCCV wall, SCCV top slab, inner mat foundation, and containment closure head.

Figure 3E-5 presents the accident pressure Pa1, selected as representative of other pressure loads applied on the containment, and the resulting deformed shape of the integrated RB due to the accident pressure.

3E.3.4 Responses and Demands from Groundwater and Rock Pressure Loads The lateral groundwater pressure is applied as an inward normal pressure with a gradient equal to the density of water at approximately 60°F (62.4 lb/ft3) multiplied by the distance starting from grade down to the centerline of the common mat foundation.

The 114.8 ft of groundwater causing buoyancy uplift pressure on the mat foundation is applied as an upward normal load on the mat foundation model.

Rock pressures described in Subsection 3E.1.3 are applied as an inward normal pressure on the embedded RB with a gradient applied circumferentially between 43.5 psi and 14.5 psi with linear interpolation in between as illustrated in Figure 3E-6. The rock pressures are applied as nodal forces on the RB exterior wall starting at a depth of 12.9 m (42.3 ft) below the grade centerline.

Responses and demands due to groundwater and rock pressure loads are consistent with the loading as demonstrated by Figure 3E-7.

3E.4 Thermal Stress Analysis Thermal stress analyses are performed by applying body loads to the shell and beam elements of the models described in Section 3B.5 to induce thermal strains. Shell elements are assigned temperatures to each face to induce a linear thermal gradient through the shell thickness while beam elements are assigned a uniform temperature throughout their length and cross-section.

Modeling requirements for thermal load cases considered for the design of the integrated RB structures are presented in Table 3B-1.

3E.4.1 Analysis Cases Table 3B-1 presents the thermal stress load cases considered for the design of the BWRX-300 integrated RB structures.

As indicated in Table 3B-1, four design basis accident conditions are considered in the heat transfer analyses in addition to the normal operating thermal conditions, with DBA-1 being a design basis accident inside containment. The heat transfer analysis only considers the extreme winter conditions with an outdoor temperature of -40oF which create the most severe thermal gradients on steel-plate composite components due to the larger differences between the outside temperature and the temperatures in the BWRX-300 containment and RB. Summer conditions are only considered for the DBA-1 analysis cases.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-8 Revision 0 Supplement 1 The FE models analyzed for normal operating thermal conditions are assigned effective structural stiffnesses, while partially cracked models are used in DBA analyses. The four DBAs result in temperatures on the walls and slabs within the DBA region which exceed the threshold temperature defined in ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, (Reference 3E-4) and require fully cracked equivalent section properties to be assigned to the models. Under normal operating temperature load conditions, the steel-plate composite elements are not further cracked beyond the effective structural stiffness because gradient loads are small and develop over significant time.

For both normal operating and DBA thermal load cases, the RB model is considered with LB and UB subgrade stiffness to determine the bounding conditions for determining internal design forces.

3E.4.2 Response and Demands from Thermal Loads The deformed shapes for normal operating and DBA winter thermal conditions are shown in Figure 3E-8.

The expansion and contraction of the RB exterior wall is consistent with the effect of the differential temperatures exterior and interior to the RB. The RPV pedestal experiences the highest expansion deformations because of the higher temperatures in the SCCV and the essentially unrestrained thermal expansion of the cantilevered component.

Internal forces from the DBA-1 thermal analysis with summer exterior temperatures and an UB subgrade condition are presented in Figure 3E-9.

3E.5 References 3E-1 ASCE/SEI 7-16, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, American Society of Civil Engineers, 2016.

3E-2 BC-TOP-3-A, Tornado and Extreme Wind Design Criteria for Nuclear Power Plants, Bechtel Power Corporation, Revision 3, August 1974.

3E-3 ASCE/SEI 7-22, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, American Society of Civil Engineers, 2022.

3E-4 ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, American National Standards Institute/American Institute of Steel Construction, 2018.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-9 Revision 0 Supplement 1 Table 3E-1 Equivalent Linear Static Subgrade Profiles Analysis Case Parameters Used 1-g SSI analysis UB, LB and UB Static and quasi-static analyses LB and LB Thermal analysis (Operational and Accident)

UB and UB LB and LB Notes:

(1)

= Effective unit weight for subgrade materials below groundwater table (2)

= Young Modulus representing linearized stiffness properties of the soil and rock for long-term static loading conditions (3)

= Poissons ratio representative of at-rest lateral earth pressure conditions

Est st Est st Est st Est st

Est st

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-10 Revision 0 Supplement 1 Table 3E-2 Crane Loading Crane load case Crane payload Load application direction Note Crane vertical payload 251.3 kip Downwards (-Z)

Crane payload capacity Crane transverse load 62.5 kip

+Y 25% of vertical payload taken as Crane transverse load Crane Hoisting impact load 41.2 kip Downwards (-Z)

Considering lifted load, load block, and attachment Crane horizontal load 62.5 kip

+X 25% of vertical payload taken as Crane horizontal load

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-11 Revision 0 Supplement 1 Table 3E-3 Containment Pressure Loading Pressure Load Pressure value kPa(g)

Psi(g)

Normal positive pressure Po+

18.7 2.7 Normal negative pressure Po-

-14

-2 Design Basis Accident 1 Peak Pressure Pa1 413.7 60 Design Basis Accident 2 Peak Pressure Pa2 839 122 Test Peak Pressure Pt 475.73 69 Isolation Condenser Pool Pressurization Pop 31.0 4.5

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-12 Revision 0 Supplement 1 Table 3E-4 Summer/Winter Thermal Conditions Inside and Outside Integrated Reactor Building During Normal Operations REGION SUMMER (or Maximum)

WINTER (or Minimum)

Main Steam/Feedwater Penetration Room 104°F 50°F Interior 104°F 50°F RB Outdoor Temperature 104°F

-40°F Isolation Condenser Pools 110°F 110°F Ground 60°F 60°F Note:

Although the governing summer ambient temperature at the CRN Site is 107°F per Table 2.0-1R, the 3oF difference is not significant as thermal stress demands from the summer condition do not govern the thermal stress demands from winter conditions which create the most severe thermal gradients on steel-plate composite components.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-13 Revision 0 Supplement 1 Table 3E-5 Breathing Mode Accelerations Floor Elevation [m(ft)]

Breathing Mode Vertical ZPA [g]

0 (0) 0.69 5.138 (16.857) 0.69 (1)

Note:

(1) For area outside the connected reactor cavity pool and equipment pool. For the reactor cavity pool and equipment pool regions, 1.76 g is used.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-14 Revision 0 Supplement 1 Figure 3E-1 Loss-of-Coolant Accident Containment Flooding

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-15 Revision 0 Supplement 1 Figure 3E-2 Internal Forces from 1-g Soil-Structure Interaction Analysis in Reactor Building Exterior Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-16 Revision 0 Supplement 1 Figure 3E-3 Internal Forces from 1-g Soil-Structure Interaction Analysis in Containment Wall

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-17 Revision 0 Supplement 1 Figure 3E-4 Deformed Shape for Hydrostatic and Hydrodynamic Loads (Units: m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-18 Revision 0 Supplement 1 Figure 3E-5 Accident Pressure Loading Inside Containment and Corresponding Deformed Shape of the Integrated Reactor Building

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-19 Revision 0 Supplement 1 Figure 3E-6 Rock Pressure on Integrated Reactor Building Exterior Wall (Units: kPa)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-20 Revision 0 Supplement 1 Figure 3E-7 Deformed Shape of Integrated Reactor Building Due to Groundwater and Rock Pressures (Units: m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-21 Revision 0 Supplement 1 Figure 3E-8 Deformed Shape of Integrated Reactor Building Due to Normal Operating Thermal Loads During Winter with Upper Bound Subgrade (Units: m)

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3E-22 Revision 0 Supplement 1 Figure 3E-9 Reactor Building Internal Forces Due to a Design Basis Accident Inside Containment during Summer

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-1 Revision 0 Supplement 1 APPENDIX 3F DESIGN DETAILS AND EVALUATION RESULTS FOR THE CONTAINMENT 3F.1 Introduction This appendix provides bounding estimates of the available margins for the SCCV structure based on results from bounding load combinations selected for the evaluation of the containment and discusses the design of critical SCCV Diaphragm Plate Steel Plate Composite (DP SC) connections.

Appendix 3F will be supplied in a Supplement in the 4th quarter of FY 2025.

Supplement 1, LCR 25-050 This appendix focuses on the design evaluation of the Steel-Plate Composite Containment Vessel (SCCV) portion of the containment and provides estimates of the available margins for the structure.

Design demands used in the evaluations are obtained from the structural analyses discussed in Appendix 3C and Appendix 3E performed using the one-step approach and the Finite Element (FE) models discussed in Appendix 3B. The design demands are obtained from design load combinations that assess the operational and accident structural demands on the structure.

3F.1 Design Methodology The design of the SCCV structure is in accordance with Section 6.0 of NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel and Reactor Building Structural Design, (Reference 3F-1).

Table 3F-1 presents the design parameters used in the DP-SC design evaluation.

3F.2 Load Combinations and Design Demands The evaluation of the SCCV is based on analysis results using load combinations as presented in Table 3.8-1.

Additional detail of the individual load demands considered is described in Subsection 3.8.1.3.

Discussion on analytical methods is captured in Subsection 3.8.1.4.

For evaluation purposes, design demands are conservatively presented on element-by-element basis without averaging of the results across adjacent elements as allowed by the NEDC-33926P.

3F.3 Structural Capacity of SCCV Diaphragm Plate Steel-Plate Composite Sections Design capacities for SCCV DP-SC components are determined using the material properties and geometric configuration of each section and are computed using the design rules for BWRX-300 containment DP-SC structures described in Section 6.0 of NEDC-33926P.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-2 Revision 0 Supplement 1 These results provide capacities for out-of-plane shear strength and the combined capacities considering out-of-plane shear forces interaction, as well as the stress capacities for steel plates and concrete infill.

The SCCV final panel section design will address how each required strength or demands, namely, axial tension, axial compression, out-of-plane flexure, and in-plane (tangential) shear compares to the corresponding allowable section strength (capacity) consistent with the acceptance criteria for all force action types specified in Table 6-1(a) and Table 6-1(b) of NEDC-33926P for factored and service load combinations.

As stated in Subsection 3.8.4.1.4, the integrated RB DP-SC faceplates are designed to accommodate the residual stress distribution due to rolling of materials during fabrication of curved sections without compromising their membrane stress capacity in both tensions and compression. An evaluation was performed to quantify the effects of residual stresses due to rolling on the curved plates within the integrated RB structures. The evaluation determined that residual stresses and strains resulting from rolling of the curved plates are minimal and do not need to be accounted for further in the design of the BWRX-300 DP-SC curved walls.

These capacities are used to calculate the demand-to-capacity ratios described in Section 3F.4.

3F.4 Structural Design Evaluation The evaluation of the SCCV structure is performed by calculating and plotting demand-to-capacity ratios for the individual loads and interaction checks discussed in Section 3F.3 for the SCCV wall, top slab, and inner mat foundation. The demand-to-capacity ratios identify critical locations and vulnerabilities for specific limit states.

Demand-to-capacity ratios are calculated for each load combination, and envelope ratios are calculated as the maximum element demand-to-capacity ratio over all the load combinations.

The enveloping ratios are used to produce an enveloped contour plot for each design criteria.

Figure 3F-1 through Figure 3F-12 present the demand-to-capacity ratios for the SCCV wall, top slab, and inner mat foundation for the selected load combinations.

These figures show the side of the SCCV wall where the main steam line penetrations are located near the top of the wall. Each set of plots show the demand-to-capacity ratios for the critical criteria for the SCCV including the principal compressive stress, faceplate Von Mises yield stress, out-of-plane shear, and out-of-plane shear interaction on the diaphragm plates.

The SCCV principal compressive stress demand-to-capacity ratios are generally less than 0.5 and the faceplate stresses are above 0.75 near the top of the SCCV with some localized exceedances around the main steam line penetrations. The out-of-plane shear plots show locally high demand-to-capacity ratios where the Reactor Building (RB) slabs and wing walls intersect the SCCV wall as is expected. Design exceedances exist in these areas, however, some of the elements with a design exceedance are within the thickness of the intersecting component or are

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-3 Revision 0 Supplement 1 outside the critical shear section. Furthermore, the finite element mesh size is approximately half the SCCV wall thickness which would allow element averaging across four elements for future evaluation.

Similar observations can be made about the SCCV top slab and inner mat foundation which shows shear-critical behavior. As noted for the SCCV wall, many of these high shear regions in the slabs lie within the intersecting walls and outside the critical shear regions.

A summary of additional evaluations performed, or design updates developed to address these exceedances will be included in the FSAR.

3F.5 Critical SCCV Connections Design The integrated RB DP-SC modules splices and floor-to-wall, wall-to-wall and wall-to-mat foundation connections are designed per Sections 5.11 and 6.14 NEDC-33926P and meet the requirements of Section N9.4 of ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, (Reference 3F-2), ANSI/AISC 360-16, Specification for Structural Steel Buildings (Reference 3F-3) and AISC Design Guide 32, Design of Modular Steel-Plate Composite Walls for Safety Related Nuclear Facilities, (Reference 3F-4).

Two types of connections are used to connect the various integrated RB steel-plate composite components as allowed by Appendix N9.4.2 of ANSI/AISC N690-18:

1.

Full-strength connections, where the required strength for the connections is 125% of the smaller of the corresponding nominal strengths of the connected parts computed per Section 5.7 of NEDC-33926P, and 2.

Overstrength connections, where the connection required strength is 200% of the required strength due to seismic loads plus 100% of the required strength due to non-seismic loads (including thermal loads).

Full-strength connections ensure a ductile behavior with yielding and inelasticity occurring away from the connection in one of the connected members. Overstrength connections are used where it is not feasible to provide a full-strength connection, such as when the sizing of the structural members is governed by non-structural requirements (e.g., shielding requirements) and the members are oversized. Demands used to design the overstrength connections are extracted from the FE model at the connection location for the applicable load combinations considered in the design and by doubling the seismic demands as required by Section N9.4.2 of ANSI/AISC N690-18.

In accordance with Section N9.4.3 of ANSI/AISC N690-18, the connections strength is calculated using the applicable force transfer mechanism and the available strength of the connectors contributing to the force transfer.

For conventional steel-plate composite module-to-module joint connections, joint shear strength is conservatively calculated per Section 21.7.4.1(c) of American Concrete Institute (ACI) 349-13 Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary (Reference 3F-5). For DP-SC module-to-module connections where the faceplates are

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-4 Revision 0 Supplement 1 continuous, or where continuity is provided through stiffener plates, such that the confinement action of the section, surrounded by the faceplates, represents a closed-filled-tube at intersections, the available joint shear capacity at intersection is calculated using Equation 5-49 of NEDC-33926P, Section 5.11, without increasing the contribution from the concrete.

The SCCV wall is connected to the inner and outer mat foundations using a rigid full-strength connection, in which the SCCV wall is the continuous member, and the inner and outer mat foundation modules are connected to the SCCV wall. This connection, referred to as a full-strength rigid T-connection, has been tested under cyclic loading with ambient and elevated temperature conditions mimicking the DBA thermal load and was found to exceed the strength and ductility acceptance criteria. See Figure 3F-13 for an illustration of these connections.

The wing walls connection to SCCV wall and floor connections to SCCV wall are identical to those connecting the wing walls and floors to the RB exterior wall discussed in Section 3H.5.

3F.6 References 3F-1 NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel (SCCV) and Reactor Building Structural Design, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, December 2024.

3F-2 ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, American National Standards Institute/American Institute of Steel Construction, 2018.

3F-3 ANSI/AISC 360-16, Specification for Structural Steel Buildings, American National Standards Institute/American Institute of Steel Construction, 2016.

3F-4 AISC Steel Design Guide 32, Design of Modular Steel-Plate Composite Walls for Safety Related Nuclear Facilities, American Institute of Steel Construction, 2017.

3F-5 ACI 349-13, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, American Concrete Institute, 2014.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-5 Revision 0 Supplement 1 Table 3F-1 Design Parameters for the SCCV Diaphragm Plate Steel-Plate Composite Modules Structural Component Member Thickness (in)

Faceplate/

Diaphragm Plate Thickness (in)

Yield Strength (ksi)

Concrete Compressive Strength (ksi)

Diaphragm Plate Spacing (in)

SCCV wall 48 0.75 60(1) 8 24 Inner mat foundation 48 0.5 8

24 SCCV top slab 60 0.75 8

30 Note:

(1) For load combinations with thermal loading, temperature-dependent material properties are used for containment.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-6 Revision 0 Supplement 1 Figure 3F-1 Containment Wall Envelope Demand-to-Capacity Ratios - Principal Compression

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-7 Revision 0 Supplement 1 Figure 3F-2 Containment Wall Envelope Demand-to-Capacity Ratios - Von Mises Stress

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-8 Revision 0 Supplement 1 Figure 3F-3 Containment Wall Envelope Demand-to-Capacity Ratios - Out-of-Plane Shear

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-9 Revision 0 Supplement 1 Figure 3F-4 Containment Wall Envelope Demand-to-Capacity Ratios - Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-10 Revision 0 Supplement 1 Figure 3F-5 Containment Top Slab Envelope Demand-to-Capacity Ratios - Principal Compression

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-11 Revision 0 Supplement 1 Figure 3F-6 Containment Top Slab Envelope Demand-to-Capacity Ratios - Von Mises Stress

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-12 Revision 0 Supplement 1 Figure 3F-7 Containment Top Slab Envelope Demand-to-Capacity Ratios - Out-of-Plane Shear

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-13 Revision 0 Supplement 1 Figure 3F-8 Containment Top Slab Envelope Demand-to-Capacity Ratios - Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-14 Revision 0 Supplement 1 Figure 3F-9 Containment Basemat (Inner Mat Foundation) Envelope Demand-to-Capacity Ratios - Principal Compression

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-15 Revision 0 Supplement 1 Figure 3F-10 Containment Basemat (Inner Mat Foundation) Envelope Demand-to-Capacity Ratios - Von Mises Stress

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-16 Revision 0 Supplement 1 Figure 3F-11 Containment Basemat (Inner Mat Foundation) Envelope Demand-to-Capacity Ratios - Out of Plane Shear

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-17 Revision 0 Supplement 1 Figure 3F-12 Containment Basemat (Inner Mat Foundation) Envelope Demand-to-Capacity Ratios - Out of Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3F-18 Revision 0 Supplement 1 Figure 3F-13 Floor to Wall Connection Location Examples

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-1 Revision 0 Supplement 1 APPENDIX 3G DESIGN DETAILS AND EVALUATION RESULTS FOR THE CONTAINMENT INTERNAL STRUCTURES 3G.1 Introduction This appendix provides bounding estimates of the available margins for the Reactor Pressure Vessel (RPV) pedestal structure based on results from bounding load combinations selected for the evaluation of the Reactor Building structure.

Appendix 3G will be supplied in a Supplement in the 4th quarter of FY 2025.

Supplement 1, LCR 25-050 This appendix focuses on the design evaluation of the Diaphragm Plate Steel-Plate Composite (DP-SC) Reactor Pressure Vessel (RPV) pedestal and provides estimates of the available margins for the structure. Design evaluations for other containment internal structures are not provided for preliminary evaluation as these structures do not contribute to the global response of the integrated RB and their design is still progressing.

Design demands used in the evaluations are obtained from the structural analyses discussed in Appendix 3C and Appendix 3E, performed using the one-step approach and the Finite Element (FE) models discussed in Appendix 3B. The design demands are obtained from design load combinations that assess the operational and accident structural demands on the structure.

3G.1 Design Methodology The design of the RPV pedestal is in accordance with Section 5.0 of NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel and Reactor Building Structural Design, (Reference 3G-1), similar to the Reactor Building (RB) structure.

Table 3G-1 presents the design parameters used in the DP-SC design evaluation.

3G.2 Load Combinations and Design Demands The evaluation of the RPV pedestal is based on design demands obtained from the load combinations in Table 3.8-7.

Additional detail of the individual load demands considered is described in Subsection 3.8.3.3.

Discussion on analytical methods is discussed in Subsection 3.8.3.4.

3G.3 Structural Capacity of Diaphragm Plate Steel-Plate Composite Sections Design capacities for the RPV pedestal DP-SC components are determined using the material properties and geometric configuration of each section and the design rules for BWRX-300 non-containment DP-SC structures described in Section 5.0 of NEDC-33926P.

These results provide capacities for uniaxial tensile strength, compressive strength, out-of-plane flexural strength, in-plane shear strength, and out-of-plane shear strength, as well as the combined capacities considering out-of-plane shear forces interaction, and in-plane membrane forces and out-of-plane moments interaction.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-2 Revision 0 Supplement 1 These capacities are used to calculate the demand-to-capacity ratios described in Section 3G.4.

3G.4 Structural Design Evaluation Demand-to-capacity ratios are calculated for each load combination, and envelope ratios are calculated as the maximum element demand-to-capacity ratio over all considered load combinations. The enveloping ratios are used to produce an enveloped contour plot for each design criterion.

Figure 3G-1 through Figure 3G-4 present the demand-to-capacity ratios for the design of the RPV pedestal for the load combinations in Table 3.8-7.

The demand-to-capacity ratios for the ANSI/AISC N690-18 (Reference 3G-2) criteria shown in Figure 3G-1 through Figure 3G-4 include the two notional half checks, out-of-plane shear, and out-of-plane shear interaction on the diaphragm plates.

Demand-to-capacity ratios for the notional half checks are approximately 0.75 near the base of the RPV pedestal. These checks include the in-plane membrane, in-plane shear, and out-of-plane bending effects. The out-of-plane shear demand-to-capacity ratios are observed to be high only at the top and bottom of the pedestal. The higher ratios at the top are due to the local interaction of the pedestal with the CEPSS structure while those at the bottom are due to the interaction with the Steel-Plate Composite Containment Vessel (SCCV) inner mat foundation.

The exceedances observed at the inner mat foundation all lie within the thickness of the SCCV mat foundation.

A summary of additional evaluations performed, or design updates developed to address these exceedances will be included in the FSAR.

3G.5 Pedestal Connection Design A full-strength rigid T-connection (see Section 3F.5 for design methodology and Figure 3F-13) is used to connect the RPV pedestal to the common mat foundation. In this connection, the inner mat foundation is the continuous member, with the RPV pedestal bearing on the mat foundation.

3G.6 References 3G-1 NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel (SCCV) and Reactor Building Structural Design, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, December 2024.

3G-2 ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, American National Standards Institute/American Institute of Steel Construction, 2018.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-3 Revision 0 Supplement 1 Table 3G-1 Design Parameters for the RPV Pedestal Diaphragm Plate Steel-Plate Composite Modules Structural Component Member Thickness (in)

Faceplate/

Diaphragm Plate Thickness (in)

Yield Strength (ksi)

Concrete Compressive Strength (ksi)

Diaphragm Plate Spacing (in)

RPV Pedestal 48 0.75 50(1) 8 24 Note:

(1) For load combinations with thermal loading, temperature-dependent material properties are used.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-4 Revision 0 Supplement 1 Figure 3G-1 Envelope Demand-to-Capacity Ratios for Reactor Pressure Vessel Pedestal - Notional Half 1

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-5 Revision 0 Supplement 1 Figure 3G-2 Envelope Demand-to-Capacity Ratios for Reactor Pressure Vessel Pedestal - Notional Half 2

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-6 Revision 0 Supplement 1 Figure 3G-3 Envelope Demand-to-Capacity Ratios for Reactor Pressure Vessel Pedestal - Out-of-Plane Shear

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3G-7 Revision 0 Supplement 1 Figure 3G-4 Envelope Demand-to-Capacity Ratios for Reactor Pressure Vessel Pedestal - Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-1 Revision 0 Supplement 1 APPENDIX 3H DESIGN DETAILS AND EVALUATION RESULTS FOR THE REACTOR BUILDING STRUCTURE 3H.1 Introduction This appendix provides bounding estimates of the available margins for key structural members of the Reactor Building based on results from bounding load combinations selected for the evaluation of the Reactor Building structure and discusses the design of critical Reactor Building DP SC connections.

Appendix 3H will be supplied in a Supplement in the 4th quarter of FY 2025.

Supplement 1, LCR 25-050 This appendix evaluates the design and provides estimates of the available margins for key structural members of the Reactor Building (RB).

Design demands used in the evaluations are obtained from the structural analyses discussed in Appendix 3C and Appendix 3E, performed using the one-step approach and the Finite Element (FE) models discussed in Appendix 3B. The design demands are obtained from design load combinations which assess the operational and accident structural demands on the structure.

3H.1 Design Methodology The design of the RB structure using Diaphragm Plate Steel-Plate Composite (DP-SC) is in accordance with Section 5.0 of NEDC-33926P, BWRX-300 Steel-Plate Composite (SC)

Containment Vessel and Reactor Building Structural Design, (Reference 3H-1) developed to address the particularities of DP-SC construction.

The design for those portions of the RB structure using conventional steel-plate composite components is in accordance with ANSI/AISC N690-18 (Reference 3H-2), Appendix N9.

Table 3H-1 presents the design parameters used in RB steel-plate composite design evaluations.

3H.2 Load Combinations and Design Demands The evaluation of the RB structure is based on results from the load combinations presented in Table 3.8-7.

Additional detail of the individual load demands considered is described in Subsection 3.8.4.1.3.

Discussion on analytical methods is captured in Subsection 3.8.4.1.4.

Similar to the Steel-Plate Composite Containment Vessel (SCCV), design demands are conservatively presented on element-by-element basis without averaging of the results across adjacent elements as allowed by the ANSI/AISC N690-18.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-2 Revision 0 Supplement 1 3H.3 Structural Capacity of RB Diaphragm Plate Steel-Plate Composite Sections Design capacities for RB steel-plate composite components are determined using the material properties and geometric configuration of each section, and are computed as follows:

DP-SC sections are evaluated using the design rules for BWRX-300 non-containment DP-SC structures described in Section 5.0 of NEDC-33926P.

Conventional steel-plate composite sections are evaluated based on ANSI/AISC N690-18 Appendix N9 as supplemented with additional guidance from AISC Steel Design Guide 32, Design of Modular Steel-Plate Composite Walls for Safety Related Nuclear Facilities, (Reference 3H-3).

These results provide capacities for uniaxial tensile strength, compressive strength, out-of-plane flexural strength, in-plane shear strength, and out-of-plane shear strength, as well as the combined capacities considering out-of-plane shear forces interaction, and in-plane membrane forces and out-of-plane moments interaction.

These capacities are used to calculate the demand-to-capacity ratios described in Section 3H.4.

3H.4 Structural Design Evaluation Demand-to-capacity ratios are calculated for each load combination, and envelope ratios are calculated as the maximum element demand-to-capacity ratio over all the load combinations.

The enveloping ratios are used to produce an enveloped contour plot for each design criteria.

Key RB components considered in the evaluation include the RB exterior wall above and below ground, the outer mat foundation, and wing walls.

Figure 3H-1 through Figure 3H-6 present the demand-to-capacity ratios for key RB components for the load combinations in Table 3.8-7.

The demand-to-capacity ratios for the RB exterior wall below grade and above grade portions are provided in Figure 3H-1 and Figure 3H-2. The side of the RB exterior wall looking from the -X to

+X direction is shown in the figures.

The RB exterior wall shows acceptable demand-to-capacity ratios for the below grade portions but locally high notional half demand-to-capacity ratios where the subgrade slabs intersect the wall. Some notional half exceedances are observed near grade where the thick grade slab intersects the RB exterior wall around the steam tunnel. The accident in the steam tunnel and pools produces high thermal and pressure demands in this area.

The demand-to-capacity ratios for the RB outer mat foundation is provided in Figure 3H-3.

Similar to the inner mat foundation, the notional half demand-to-capacity ratios of the outer mat foundation are quite low, with high out-of-plane shear demand-to-capacity ratios near the intersections with the walls above.

The demand-to-capacity ratios for a set of the RB wing walls are provided in Figure 3H-4. These represent the wing walls that lie within the second quadrant (30 degree clockwise from -X to +Y) and the fourth quadrant (30 degree clockwise from +X to -Y). The notional half

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-3 Revision 0 Supplement 1 demand-to-capacity ratios show values generally between 0.5 and 0.75. Some locally high out-of-plane shear is observed near the top where the wing walls are attached to the thick slabs near grade.

The RB subgrade floor at EL-14.8m (-48.6 ft) and the refueling floor are provided in Figure 3H-5 and Figure 3H-6. The demand-to-capacity ratios for the subgrade floor are all quite low as these floors have been thickened to increase their vertical stiffness and natural frequency for vibratory purposes. The refueling floor has generally acceptable demand-to-capacity ratios with some exceedances around the south side where the thick fuel pool walls below intersect.

A summary of additional evaluations performed. or design updates developed to address these exceedances will be included in the FSAR.

3H.5 Critical Reactor Building Connections Design The RB exterior wall-to-outer mat foundation connection is a full-strength rigid connection, in which the RB wall is the continuous member, with the mat foundation connected to the wall. This connection is referred as a full-strength rigid L-connection.

Wing walls, and slab at grade-to-RB and SCCV walls, are also connected using full-strength rigid T-connections, while below-grade slabs are connected to the RB and SCCV walls using overstrength rigid T-connections for out-of-plane moment transfer.

The RB roof-to-RB exterior wall is a full-strength L-connection. In this connection, the RB reinforced concrete roof is the continuous member and bearing on the RB exterior wall. This connection is classified as rigid connection for out-of-plane moment transfer due to the continuity of the concrete and headed dowels used to transfer the out-of-plane moment.

For the design methodology of these connections, refer to Section 3F.5. See Figure 3F-13 for an example of these connections.

3H.6 References 3H-1 NEDC-33926P, BWRX-300 Steel-Plate Composite (SC) Containment Vessel (SCCV) and Reactor Building Structural Design, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, December 2024.

3H-2 ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, American National Standards Institute/American Institute of Steel Construction, 2018.

3H-3 AISC Steel Design Guide 32, Design of Modular Steel-Plate Composite Walls for Safety Related Nuclear Facilities, American Institute of Steel Construction, 2017.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-4 Revision 0 Supplement 1 Table 3H-1 Design Parameters for the Reactor Building Steel-Plate Composite Components Structural Component Component Type Member Thickness (in)

Faceplate Thickness (in)

Yield Strength (ksi)

Concrete Compressive Strength (ksi)

Diaphragm Plate Spacing (in)

RB exterior wall (elevation -5 m

(-16.4 ft) and lower)

DP-SC 36 0.75 65 8

18 RB exterior wall (elevation -5 m

(-16.4 ft) to 0.0 m (0.0 ft))

DP-SC 36 1.25 18 RB exterior wall (elevation 0.0 m (0.0 ft) to 18.3 m (60.0 ft))

DP-SC 36 1.25 18 RB exterior wall (elevation 18.3 m (60.0 ft) and above)

DP-SC 36 0.75 18 Outer mat foundation DP-SC 48 0.5 24 Wing Walls Conventional 18 0.5 50 9(1)

Steam Tunnel (Thick Wing) Walls DP-SC 36 0.875 50 18 Note:

(1) Tie bar spacing provided vs the diaphragm spacing provided for DP-SC components.

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-5 Revision 0 Supplement 1 Figure 3H-1 Envelope Demand-to-Capacity Ratios for Reactor Building Exterior Wall Below Grade (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-6 Revision 0 Supplement 1 Figure 3H-2 Envelope Demand-to-Capacity Ratios for the Reactor Building Exterior Wall Above Grade (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-7 Revision 0 Supplement 1 Figure 3H-3 Outer Mat Foundation Envelope Demand-to-Capacity Ratios (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-8 Revision 0 Supplement 1 Figure 3H-4 Envelope Demand-to-Capacity Ratios for Reactor Building Wing Walls (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-9 Revision 0 Supplement 1 Figure 3H-5 Envelope Demand-to-Capacity Ratios for Reactor Building Subgrade Floor EL-14.8m (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction

Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3H-10 Revision 0 Supplement 1 Figure 3H-6 Envelope Demand-to-Capacity Ratios for Reactor Building Refuel Floor (a) Notional Half 1, (b) Notional Half 2, (c) Out-of-Plane Shear, and (d) Out-of-Plane Shear Interaction