ML18260A105

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LLC Supplemental Response to NRC Request for Additional Information No. 151 (Erai No. 8974) on the NuScale Design Certification Application
ML18260A105
Person / Time
Site: NuScale
Issue date: 09/17/2018
From: Rad Z
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-0918-61804
Download: ML18260A105 (46)
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RAIO-0918-61804 Docket No.52-048 September 17, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Supplemental Response to NRC Request for Additional Information No. 151 (eRAI No. 8974) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 151 (eRAI No. 8974)," dated August 05, 2017
2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 151 (eRAI No.8974)," dated October 03, 2017
3. NuScale Power, LLC Supplemental Response to NRC "Request for Additional Information No. 151 (eRAI No.8974)," dated April 09, 2018 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) supplemental response to the referenced NRC Request for Additional Information (RAI).

The Enclosure to this letter contains NuScale's supplemental response to the following RAI Question from NRC eRAI No. 8974:

  • 03.08.04-23 This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

Sincerely,

---: ? ~

~~

Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A Marieliz Vera, NRC, OWFN-8G9A Enclosure 1: NuScale Supplemental Response to NRC Request for Additional Information eRAI No.8974 NuScale Power, LLC 1100 NE Circle Blvd. , Suite 200 Corvalis, Oregon 97330 , Office: 541.360.0500 , Fax: 541.207.3928 www.nuscalepower.com

RAIO-0918-61804 :

NuScale Supplemental Response to NRC Request for Additional Information eRAI No. 8974 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 8974 Date of RAI Issue: 08/05/2017 NRC Question No.: 03.08.04-23 10 CFR 50, Appendix A, GDC 1, 2, and 4, provide requirements to be met by SSC important to safety. In accordance with these requirements, DSRS Sections 3.7.1 and 3.8.4 provide review guidance pertaining to seismic parameters and design of seismic Category I structures.

Consistent with the guidance in DSRS 3.7.1.II.4.A.viii, the staff reviews comparison criteria for the acceptability of a standard design for a potential site.

COL item 3.8-2 in Section 3.8.4.8 directs the COL applicant to confirm that the site independent RXB and CRB are acceptable for use at the designated site. Further, Section 3.8.4.8 identifies locations within the building and respective ISRS which are to be used by the COL applicant to compare with their respective site-specific ISRS for purposes of confirming the acceptability of the site independent structures for the designated site. The applicant is requested to correct inconsistencies between the ISRS Figures referred to in FSAR Section 3.8.4.8 and the respective Figures in FSAR Section 3.7. Further, clarify whether the ISRS in these figures are based on the envelope of all or a partial envelope of the SSI and SSSI analysis cases.

Further, the staff request the applicant to address the following in the FSAR.

1. propose locations for the comparison of building member forces and deformations, with the identification of the respective FSAR Tables and Figures
2. clarify whether the current locations for ISRS comparison include responses at peripheral locations to detect rocking and torsion or propose additional locations as necessary
3. augment the list of locations for ISRS comparison in the RXB to address the fuel racks
4. include responses to check overturning, torsional, and sliding stability of the structures NuScale Nonproprietary

NuScale Response:

As discussed in an NRC Public meeting on June 12, 2018, the NRC provided supplemental questions to eRAI 8974 Question 03.08.04-23 as follows:

1. The staff requests the applicant to also indicate in COL Items 3.7-5 ISRS as part of the seismic demands to be investigated by the COL applicant.

Response: COL Item 3.7-5 has been revised to state the following:

"A COL applicant that references the NuScale Power Plant design certification will perform a soil-structure interaction analysis of the Reactor Building and the Control Building using the NuScale SASSI2010 models for those structures. The COL applicant will confirm that the site-specific seismic demands of the standard design for critical structures, systems, and components in Appendix 3B are bounded by the corresponding design certified seismic demands and, if not, the standard design for critical structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. Seismic demands investigated shall include forces, moments, deformations, in-structure response spectra, and seismic stability of the structures."

2. The staff requests the applicant to provide FSAR markups identifying the specific FSAR Figures and/or Tables and/or Sections containing the standard design forces, moments, deformations, and seismic stability of the structures to be used in the comparison, also including identification of the corresponding locations throughout the structures and identification of the node(s) and or element(s) comprising the individual and/or envelop responses of ISRS, forces and moments, deformations, and seismic stability results, as applicable, consistent with RAI 8935, Q25.

Response: This staff request will be responded to in RAI 8935 Question 03.07.02-25.

For ISRS, Table 3.7.2-53 is added to the FSAR to list the nodes enveloped at each floor to produce the floor ISRS. Figures 3.7.2-142 through 3.7.2-148 are added to the FSAR to show the locations of the nodes.

3. The staff request for ISRS at wall locations is for ISRS that include results from nodes at or near the center of the wall so as to capture the amplified out-of-plane response of the wall.

Therefore, the staff requests the applicant to provide ISRS at wall locations that include the results from nodes at or near the center of the wall and address such ISRS also under COL Item 3.8-2.

NuScale Nonproprietary

Response: There is no safety-related equipment attached to the exterior walls in the NuScale design, and, thus, no center-of-wall ISRS will be presented. If equipment moves to the exterior wall, this will be re-evaluated, and appropriate spectra will be developed and used to analyze that equipment.

4. Further, the staff requests the applicant to explain how the floor ISRS are developed; whether they account for contributions from all or selected nodes on that floor (if selected, criteria and justification for such selection) and a method of incorporation (averaging or enveloping).

Response: The horizontal and vertical ISRS at selected locations of the RXB are calculated at 2, 3, 4, 5, 7, and 10% damping ratios. In the SASSI analysis, the analysis model is subjected to only one component of seismic excitation at a time for each soil type and one of the two concrete conditions (i.e., cracked or uncracked).

The procedure for generating the ISRS follows the guidance of RG 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components", Rev. 1 (also see FSAR Section 3.7.2.5):

Step 1. Obtain the three acceleration response time histories at each selected nodal location due to the EW, NS, and vertical components of a seismic input.

Step 2. Calculate the ISRS of each response time history at 196 discrete frequencies provided in Table 1. These 196 frequencies include all frequencies given in the NRC Standard Review Plan (SRP) Section 3.7.1, Table 3.7.1-1 and an intermediate frequency between any two consecutive frequencies specified in Table 3.7.1-1 and some additional frequencies below 0.2 Hz.

NuScale Nonproprietary

Table 1 - The 196 Frequencies for ISRS Generation.

Row 196 Frequencies (Hz) for ISRS Generation Listed Row-wise in Increasing Frequencies.

No.

1 0.1 0.1125 0.125 0.1375 0.15 0.1625 0.175 0.1875 0.2 0.225 2 0.25 0.275 0.3 0.325 0.35 0.375 0.4 0.425 0.45 0.475 3 0.5 0.525 0.55 0.575 0.6 0.65 0.7 0.75 0.8 0.85 4 0.9 0.95 1 1.05 1.1 1.15 1.2 1.25 1.3 1.35 5 1.4 1.45 1.5 1.55 1.6 1.65 1.7 1.75 1.8 1.85 6 1.9 1.95 2 2.05 2.1 2.15 2.2 2.25 2.3 2.35 7 2.4 2.45 2.5 2.55 2.6 2.65 2.7 2.75 2.8 2.85 8 2.9 2.95 3 3.075 3.15 3.225 3.3 3.375 3.45 3.525 9 3.6 3.7 3.8 3.9 4 4.1 4.2 4.3 4.4 4.5 10 4.6 4.7 4.8 4.9 5 5.125 5.25 5.375 5.5 5.625 11 5.75 5.875 6 6.125 6.25 6.375 6.5 6.625 6.75 6.875 12 7 7.125 7.25 7.375 7.5 7.625 7.75 7.875 8 8.25 13 8.5 8.75 9 9.25 9.5 9.75 10 10.25 10.5 10.75 14 11 11.25 11.5 11.75 12 12.25 12.5 12.75 13 13.25 15 13.5 13.75 14 14.25 14.5 14.75 15 15.5 16 16.5 16 17 17.5 18 19 20 21 22 23.5 25 26.5 17 28 29.5 31 32.5 34 35.5 37 38.5 40 41.5 18 43 44.5 46 47.5 49 50.5 52 53.5 55 56.5 19 58 59.5 61 64 67 70 73 76 79 82 20 85 88 91 94 97 100 - - - -

Step 3. At each selected location, combine the three co-directional ISRS due to the three components of the seismic input by the SRSS method.

Step 4. Repeat Steps 1 through 3 for each seismic input.

Step 5. Average the ISRS that are obtained in Step 4 due to five CSDRS-compatible inputs.

Note that, for CSDRS-HF-compatible input, no average is necessary because there is only one CSDRS-HF-compatible input.

Step 6. Repeat Steps 1 through 5 for different soil types.

Step 7. Repeat Steps 1 through 6 for the cracked and uncracked concrete conditions.

Step 8. The ISRS are enveloped over soil types and concrete conditions considered.

Step 9. Broaden each enveloped ISRS by +/-15% to account for structural model uncertainties in accordance with ASCE 4-98 requirements. The broadening is identical to that specified in NRC Regulatory Guide 1.122. The enveloped ISRS is broadened 15% on the frequency scale; i.e., a frequency band of 0.85xfi to 1.15xfi is used to widen the spectral acceleration of the ith spectral frequency fi.

NuScale Nonproprietary

Step 10. Envelope single and triple building ISRS for each selected node. The nodes selected for ISRS generation were chosen because they would provide maximum or representative responses. For example, at EL 24' (top of foundation in gallery), the northwest and northeast corner nodes of the RXB outer wall experience the highest displacements and, therefore, ISRS. Another approximate mid-point on the outer wall (gridline 4) was selected to provide an intermediate ISRS. At EL 25', (top of pool floor), the 6 nodes selected were at the center base node of NPM1 through NPM6. These 6 nodes lie on the north side of the RXB.

Because the RXB is symmetric about the east-west axis, the ISRS beneath the NPMs on the south side will be nearly identical.

Step 11: Floor ISRS The ISRS at select locations on the same floor are enveloped to obtain the floor ISRS.

Table 3.7.2-53 is added to the FSAR to list the nodes enveloped at each floor to produce the floor ISRS. Figures 3.7.2-142 through 3.7.2-148 have been added to the FSAR to show the locations of the nodes.

5. The staff requests the applicant to provide an ISRS at the NPM support skirt location that is consistent with the input used for the NPM analysis documented in TR-0916-51502 and identify the node(s) for such ISRS.

Response: Please see FSAR markup.

At the CNV skirts of NPM1 and NPM6, response spectra are generated for the time histories at the eight spider nodes, corresponding to the eight spider elements. The SASSI node numbers are listed in Table 2 of this response.

The resulting spider node spectra are then averaged for each module. This results in nine averaged skirt response spectra for each module, based on the three seismic cases provided (Soil Type 7, Capitola time history, cracked concrete nominal stiffness, cracked concrete reduced stiffness, uncracked concrete nominal stiffness), each with three components (X,Y, and Z). The ISRS of the nine averaged skirt response spectra is then enveloped for NPM1 and NPM6 in the X, Y, and Z directions. The six resulting enveloping ISRS (two modules x one skirt support x three directions) for the NPM1 and NPM6 CNV skirts are shown in Figures 1 and 2.

NuScale Nonproprietary

Table 2. SASSI CNV skirt nodes NPM1 NPM6 6027 6287 6028 6288 6029 6289 6039 6299 6042 6302 6053 6307 6054 6308 6055 6309 At the CNV lugs of NPM1 and NPM6, response spectra are generated for the time histories at the nodes listed in Table 3. The spectra are then enveloped at each of the lugs on NPM1 and NPM6, resulting in 18 total enveloping spectra (two modules x three lugs x three directions). These spectra are shown in Figures 3 through 8.

Table 3. SASSI CNV lug nodes NPM1 NPM6 West Lug 6477 31081 North Lug 6483 31087 East Lug 6486 31090 NuScale Nonproprietary

Figure 1 - Enveloping ISRS of cases 1, 2 and 3 at the CNV skirt of NPM1 NuScale Nonproprietary

Figure 2 - Enveloping ISRS of cases 1,2 and 3 at the CNV skirt of NPM6 NuScale Nonproprietary

Figure 3 - Enveloping ISRS of cases 1,2 and 3 at the East Lug of NPM1 NuScale Nonproprietary

Figure 4 - Enveloping ISRS of cases 1,2 and 3 at the North Lug of NPM1 NuScale Nonproprietary

Figure 5 - Enveloping ISRS of cases 1,2 and 3 at the West Lug of NPM1 NuScale Nonproprietary

Figure 6 - Enveloping ISRS of cases 1,2 and 3 at the East Lug of NPM6 NuScale Nonproprietary

Figure 7 - Enveloping ISRS of cases 1,2 and 3 at the North Lug of NPM6 NuScale Nonproprietary

Figure 8 - Enveloping ISRS of cases 1,2 and 3 at the West Lug of NPM6 NuScale Nonproprietary

6. The staff requests the applicant to provide FSAR markups identifying the specific FSAR Figures and/or Tables and/or Sections containing the standard design responses for Items 1),

2), and 4) of COL Item 3.7-10.

a. ISRS of standard design at foundation and roof.

Response: Foundation ISRS can be found in FSAR Figures 3.7.2-107 and 3.7.2-108. The roof ISRS can be found in Figure 3.7.2-113.

b. Maximum forces in NPM lug restraints and skirts.

Response: NuScale provided the maximum forces in NPM lug restraints and skirts in RAI 8936 Question 03.07.02-10.

c. Max forces and moments in the east and west wing walls and pool walls.

Response: Max forces and moments in the east and west wing walls and pool walls can be found in FSAR Table 3.7.2-32.

Impact on DCA:

FSAR Tier 2, Section 3.7.2 has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Final Safety Analysis Report Interfaces with Certified Design RAI 01-61, RAI 02.04.13-1, RAI 03.04.01-4, RAI 03.04.02-1, RAI 03.04.02-2, RAI 03.04.02-3, RAI 03.05.01.04-1, RAI 03.05.02-2, RAI 03.06.02-15, RAI 03.06.03-11, RAI 03.07.01-2, RAI 03.07.01-3, RAI 03.07.02-8, RAI 03.07.02-12, RAI 03.08.04-23S1, RAI 03.08.04-23S2, RAI 03.08.05-14S1, RAI 03.09.02-15, RAI 03.09.02-48, RAI 03.09.02-67, RAI 03.09.02-69, RAI 03.09.03-12, RAI 03.09.06-5, RAI 03.09.06-6, RAI 03.09.06-16, RAI 03.09.06-16S1, RAI 03.09.06-27, RAI 03.11-8, RAI 03.11-14, RAI 03.11-14S1, RAI 03.11-18, RAI 03.13-3, RAI 04.02-1S2, RAI 05.02.03-19, RAI 05.02.05-8, RAI 05.04.02.01-13, RAI 05.04.02.01-14, RAI 06.02.06-22, RAI 06.02.06-23, RAI 06.04-1, RAI 09.01.02-4, RAI 09.01.05-3, RAI 09.01.05-6, RAI 09.03.02-3, RAI 09.03.02-4, RAI 09.03.02-5, RAI 09.03.02-6, RAI 09.03.02-8, RAI 10.02-1, RAI 10.02-2, RAI 10.02-3, RAI 10.02.03-1, RAI 10.02.03-2, RAI 10.03.06-1, RAI 10.03.06-5, RAI 10.04.06-1, RAI 10.04.06-2, RAI 10.04.06-3, RAI 10.04.10-2, RAI 11.01-2, RAI 13.01.01-1, RAI 13.01.01-1S1, RAI 13.02.02-1, RAI 13.03-4, RAI 13.05.02.01-2, RAI 13.05.02.01-2S1, RAI 13.05.02.01-3, RAI 13.05.02.01-3S1, RAI 13.05.02.01-4, RAI 13.05.02.01-4S1, RAI 14.02-7, RAI 19-31, RAI 19-31S1, RAI 19-38, RAI 20.01-13 Table 1.8-2: Combined License Information Items Item No. Description of COL Information Item Section COL Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.1 site-specific plant location.

COL Item 1.1-2: A COL applicant that references the NuScale Power Plant design certification will provide the 1.1 schedules for completion of construction and commercial operation of each power module.

COL Item 1.4-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.4 prime agents or contractors for the construction and operation of the nuclear power plant.

COL Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site- 1.7 specific diagrams and legends, as applicable.

COL Item 1.7-2: A COL applicant that references the NuScale Power Plant design certification will list additional 1.7 site-specific piping and instrumentation diagrams and legends as applicable.

COL Item 1.8-1: A COL applicant that references the NuScale Power Plant design certification will provide a list of 1.8 departures from the certified design.

COL Item 1.9-1: A COL applicant that references the NuScale Power Plant design certification will review and 1.9 address the conformance with regulatory criteria in effect six months before the docket date of the COL application for the site-specific portions and operational aspects of the facility design.

COL Item 1.10-1: A COL applicant that references the NuScale Power Plant design certification will evaluate the 1.10 potential hazards resulting from construction activities of the new NuScale facility to the safety-related and risk significant structures, systems, and components of existing operating unit(s) and newly constructed operating unit(s) at the co-located site per 10 CFR 52.79(a)(31). The evaluation will include identification of management and administrative controls necessary to eliminate or mitigate the consequences of potential hazards and demonstration that the limiting conditions for operation of an operating unit would not be exceeded. This COL item is not applicable for construction activities (build-out of the facility) at an individual NuScale Power Plant with operating NuScale Power Modules.

COL Item 2.0-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate 2.0 that site-specific characteristics are bounded by the design parameters specified in Table 2.0-1.

If site-specific values are not bounded by the values in Table 2.0-1, the COL applicant will demonstrate the acceptability of the site-specific values in the appropriate sections of its combined license application.

COL Item 2.1-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.1 site geographic and demographic characteristics.

COL Item 2.2-1: A COL applicant that references the NuScale Power Plant design certification will describe 2.2 nearby industrial, transportation, and military facilities. The COL applicant will demonstrate that the design is acceptable for each potential accident, or provide site-specific design alternatives.

COL Item 2.3-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.3 site-specific meteorological characteristics for Section 2.3.1 through Section 2.3.5, as applicable.

COL Item 2.4-1: A COL applicant that references the NuScale Power Plant design certification will investigate 2.4 and describe the site-specific hydrologic characteristics for Section 2.4.1 through Section 2.4.14, as applicableexcept Section 2.4.8 and Section 2.4.10.

COL Item 2.5-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.5 site-specific geology, seismology, and geotechnical characteristics for Section 2.5.1 through Section 2.5.5, below.

Tier 2 1.8-3 Draft Revision 2

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued)

Item No. Description of COL Information Item Section COL Item 3.6-2: A COL applicant that references the NuScale Power Plant design certification will verify that the 3.6 pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines in the reactor pool bay is applicable. If changes are required, the COL applicant will update the pipe rupture hazards analysis, design additional protection features as necessary, and update Table 3.6-2, Figure 3.6-12, Figure 3.6-13, Figure 3.6-14, and Figure 3.6-15 as appropriate.

COL Item 3.6-3: A COL applicant that references the NuScale Power Plant design certification will perform the 3.6 pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the reactor pool bay and design appropriate protection features.

This includes an evaluation and disposition of multi-module impacts in common pipe galleries, the identification of any new detection and auto-isolation functions for mitigating an auxiliary boiler high-energy line break, and evaluations regarding subcompartment pressurization. The COL applicant will update Table 3.6-2, Figure 3.6-16, and Figure 3.6-17 as appropriate.

COL Item 3.6-4: Not used. 3.6 COL Item 3.7-1: A COL applicant that references the NuScale Power Plant design certification will describe the 3.7 site-specific structures, systems, and components.

COL Item 3.7-2: A COL applicant that references the NuScale Power Plant design certification will provide site- 3.7 specific time histories. In addition to the above criteria for cross correlation coefficients, time step and earthquake duration, strong motion durations, comparison to response spectra and power spectra density, the applicant will also confirm that site-specific ratios V/A and AD/V2 (A, V, D, are peak ground acceleration, ground velocity, and ground displacement, respectively) are consistent with characteristic values for the magnitude and distance of the appropriate controlling events defining the site-specific uniform hazard response spectra.

COL Item 3.7-3: A COL applicant that references the NuScale Power Plant design certification will: 3.7

  • develop a site-specific strain compatible soil profile.
  • confirm that the criterion for the minimum required response spectrum has been satisfied.
  • determine whether the seismic site characteristics fall within the seismic design parameters such as soil layering assumptions used in the certified design, range of soil parameters, shear wave velocity values, and minimum soil bearing capacity.

COL Item 3.7-4: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.7 nearby structures exposed to a site-specific safe shutdown earthquake will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building.

COL Item 3.7-5: A COL applicant that references the NuScale Power Plant design certification will perform a soil- 3.7 structure interaction analysis of the Reactor Building and the Control Building using the NuScale SASSI2010 models for those structures. The COL applicant will confirm that the site-specific seismic demands of the standard design for critical structures, systems, and components in Appendix 3B are bounded by the corresponding design certified seismic demands and, if not, the standard design for critical structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. Seismic demands investigated shall include forces, moments, deformations, in-structure response spectra, and seismic stability of the structures.

COL Item 3.7-6: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 structure-soil-structure interaction analysis that includes the Reactor Building, Control Building, Radioactive Waste Building and both Turbine Generator Buildings. The COL applicant will confirm that the site-specific seismic demands of the standard design structures, systems, and components are bounded by the corresponding design certified seismic demands and, if not, the standard design structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands.

COL Item 3.7-7: A COL applicant that references the NuScale Power Plant design certification will provide a 3.7 seismic monitoring system and a seismic monitoring program that satisfies Regulatory Guide 1.12 Nuclear Power Plant Instrumentation for Earthquakes, Rev. 2 (or later) and Regulatory Guide 1.166 Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-earthquake Actions, Rev. 0 (or later). This information is to be provided as noted below.

Tier 2 1.8-5 Draft Revision 2

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued)

Item No. Description of COL Information Item Section COL Item 3.7-8: A COL applicant that references the NuScale Power Plant design certification will identify the 3.7 implementation milestone for the seismic monitoring program. In addition, a COL applicant that references the NuScale Power Plant design certification will prepare site-specific procedures for activities following an earthquake. These procedures and the data from the seismic instrumentation system will provide sufficient information to determine if the level of earthquake ground motion requiring shutdown has been exceeded. An activity of the procedures will be to address measurement of the post-seismic event gaps between the fuel racks and the pool walls and between the individual fuel racks and to take appropriate corrective action if needed (such as repositioning the racks or assuring that the as-found condition of the racks is acceptable based on the assumptions of the racks' design basis analysis). Acceptable guidance for procedure development is contained in Regulatory Guide 1.166 "Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-earthquake Actions," Rev. 0 (or later) and 1.167, "Restart of a Nuclear Power Plant Shut Down by a Seismic Event," Rev. 0 (or later).

COL Item 3.7-9: A COL applicant that references the NuScale Power Plant design certification will include an 3.7 analysis of performance-based response spectra established at the surface and intermediate depth(s) that take into account the complexities of the subsurface layer profiles of the site and provide a technical justification for the adequacy of V/H spectral ratios used in establishing the site-specific foundation input response spectra and performance-based response spectra for the vertical direction.

COL Item 3.7-10: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 site-specific configuration analysis that includes the Reactor Building with applicable configuration layout of the desired NuScale Power Modules. The COL applicant will confirm the following are bounded by the corresponding design certified seismic demands:

1) The in-structure response spectra of the standard design at the foundation and roof. See FSAR Figure 3.7.2-107 and Figure 3.7.2-108 for foundation in-structure response spectra and Figure 3.7.2-113 for roof in-structure response spectra.
2) The maximum forces in the NuScale Power Module lug restraints and skirts.
3) The site-specific in-structure response spectra for the NuScale Power Module at the skirt support will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-156 and Figure 3.7.2-157. The site-specific in-structure response spectra for the NuScale Power Module at the lug restraints will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-158 through Figure 3.7.2-163.
4) The maximum forces and moments in the east and west wing walls and pool walls. See FSAR Table 3.7.2-32.
5) The site-specific in-structure response spectra for the fuel storage racks will be shown to be bounded by the in-structure response spectra in Figure 3-6 through Figure 3-14 of TR-0816-49833.
6) The site-specific in-structure response spectra shown immediately below will be shown to be bounded by their corresponding certified in-structure response spectra:
  • Reactor Building north exterior wall at EL 75-0: bounded by in-structure response spectra in Figure 3.7.2-110
  • Reactor Building west exterior wall at EL 126-0: bounded by in-structure response spectra in Figure 3.7.2-112
  • Reactor Building crane wheels at EL 145-6: bounded by in-structure response spectra in Figure 3.7.2-114
  • Control Building east wall at EL 76-6: bounded by in-structure response spectra in Figure 3.7.2-119
  • Control Building south wall at EL 120-0: bounded by in-structure response spectra in Figure 3.7.2-121 If not, the standard design will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands.

Tier 2 1.8-6 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Step 4. For each selected area, all of the ISRS (this usually includes more than one node) are combined and the envelope obtained for each of the three directions.

Step 5. Each envelope response spectra is broadened by +/-15%.

Step 6. Steps 1 through 5 are repeated to generate ISRS at damping ratios of 2%, 3%, 4%, 5%, 7%, and 10%.

This process is shown for a single node in Figure 3.7.2-99 through Figure 3.7.2-103.

The first three figures show the development of the average ISRS for the three soil cases (7, 8, and 9) and two stiffnesses (cracked and uncracked). Figure 3.7.2-102 shows the combination of averages and the development of the ISRS envelope.

The upper three plots show this process for the CSDRS compatible time histories and soil cases and the bottom three plots show the process for the ISRS from the CSDRS-HF compatible time histories and soil cases. Figure 3.7.2-103 shows the development of the broadened spectra at various damping values. The upper three plots show the envelop ISRS for each direction and the different damping ratios. In these plots the broadening of the 2 percent damping results is shown. The bottom three plots provide the broadened results for all damping ratios.

3.7.2.5.2 Comparison of In-Structure Response Spectra between Single and Triple Building Models The structure-soil-structure interaction of the triple model has an effect on the ISRS of the RXB. Other than the ISRS at top of basemat, the ISRS of the standalone model are higher than those of the triple building model. The reduction in the ISRS of the triple building model is attributed to the extra damping effect provided by the close presence of the RWB and the CRB on the sides of the RXB.

This can be seen in Figure 3.7.2-104, Figure 3.7.2-105 and Figure 3.7.2-106.

The ISRS from the triple building model were not created for the CRB.

Because neither the standalone nor triple building model produce bounding results at all locations, ISRS enveloping the two models are used for design of structures, systems, and components in the RXB.

3.7.2.5.3 Reactor Building In-Structure Response Spectra RAI 03.08.04-23, RAI 03.08.04-23S2 For convenience in design of components and supports that need to be Seismic Category I or Seismic Category II, ISRS at multiple nodes at each floor are combined to develop a single ISRS for each floor. The ISRS corresponding to each main floor of the RXB identified below are provided oin the listed figures. Although ISRS are provided at the NPM base (floor at EL. 25' 0"), time histories were used as input for the evaluation of the NPMs as described in Appendix 3A. The governing ISRS envelop the ISRS taken from node locations on the corners of the buildings to capture the torsional and rocking components. See Table 3.7.2-53 for a list of nodes Tier 2 3.7-136 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design enveloped at each floor to produce the floor ISRS. Figure 3.7.2-142 through Figure 3.7.2-148 show the locations of the nodes selected for floor ISRS generation.

Floor Figure 24-0 Figure 3.7.2-107 25-0 Figure 3.7.2-108 50-0 Figure 3.7.2-109 75-0 Figure 3.7.2-110 100-0 Figure 3.7.2-111 126-0 Figure 3.7.2-112 181-0 Figure 3.7.2-113 3.7.2.5.4 Reactor Building Crane In-Structure Response Spectra The seismic analysis of the RBC uses ISRS for input. ISRS are generated at four selected individual crane wheel locations. These locations are on the reactor pool wall at the crane rail slab at elevation 145' 6. The enveloping ISRS for these four locations are provide in Figure 3.7.2-114. The seismic analysis of the RBC is completed per ASME NOG-1 (Reference 3.7.2-4).

3.7.2.5.5 NuScale Power Module Bay Wall In-Structure Response Spectra The NPM lug restraints transfer the horizontal forces between the NPMs and the walls of the bay. These restraints ensure that the NPM will not fall as a result of a seismic event. Because of the significance of the restraints, bounding ISRS are provided. These ISRS are not used for the design of any of the restraints, nor any specific components. However they are used in Section 3.8.4.8 to confirm acceptability of the site independent Reactor Building for use at specific sites.

Figure 3.7.2-96 provides node locations that were used to develop the ISRS for the NPM bay walls at the pool floor. The enveloping ISRS for these locations are provided in Figure 3.7.2-115.

Figure 3.7.2-97 provides node locations that were used to develop the ISRS for the NPM bay walls at the lug restraints. The enveloping ISRS for these locations are provided in Figure 3.7.2-116.

3.7.2.5.6 Control Building In-Structure Response Spectra RAI 03.08.04-23 The ISRS corresponding to each main floor of the CRB identified below are provided on the listed figures. The governing ISRS envelop the ISRS taken from node locations on the corners of the buildings to capture the torsional and rocking components.

Tier 2 3.7-137 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-97 provides node locations that were used to develop the ISRS for the NPM bay walls at the lug restraints. The enveloping ISRS for these locations are provided in Figure 3.7.2-116.

RAI 03.08.04-23S2 3.7.2.5.6 NuScale Power Module Skirt and Lug Supports ISRS RAI 03.08.04-23S2 At the CNV skirts of NPM1 and NPM6, response spectra are generated for the time histories at the eight spider nodes, corresponding to the eight spider elements. The SASSI node numbers are listed in Table 3.7.2-54.

RAI 03.08.04-23S2 The resulting spider node spectra are then averaged for each module. This results in nine averaged skirt response spectra for each module, based on the three seismic cases provided (Soil Type 7, Capitola time history, cracked concrete nominal stiffness, cracked concrete reduced stiffness, uncracked concrete nominal stiffness), each with three components (X,Y, and Z). The ISRS of the nine averaged skirt response spectra is then enveloped for NPM1 and NPM6 in the X, Y, and Z directions. The six resulting enveloping ISRS (two modules x one skirt support x three directions) for the NPM1 and NPM6 CNV skirts are shown in Figure 3.7.2-156 and Figure 3.7.2-157.

RAI 03.08.04-23S2 At the CNV lugs of NPM1 and NPM6, response spectra are generated for the time histories at the nodes listed in Table 3.7.2-55. The spectra are then enveloped at each of the lugs on NPM1 and NPM6, resulting in 18 total enveloping spectra (two modules x three lugs x three directions). These spectra are shown in Figure 3.7.2-158 through Figure 3.7.2-163.

3.7.2.5.7 Control Building In-Structure Response Spectra RAI 03.08.04-23 The ISRS corresponding to each main floor of the CRB identified below are provided on the listed figures. The governing ISRS envelop the ISRS taken from node locations on the corners of the buildings to capture the torsional and rocking components.

Floor Figure 50-0 Figure 3.7.2-117 63-3 Figure 3.7.2-118 76-6 Figure 3.7.2-119 100-0 Figure 3.7.2-120 Tier 2 3.7-142 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design 3.7.2.14 Determination of Dynamic Stability of Seismic Category I Structures Section 3.8.5 provides discussion regarding bearing pressure, lateral wall pressure, overturning, sliding, and flotation.

3.7.2.15 Analysis Procedure for Damping RAI 03.07.02-11 Section 3.7.1.2 describes the damping ratios used for seismic analysis of the RXB and CRB. As stated in Section 3.7.1.2.1, for analyses of Seismic Category I SSC, the damping values of RG 1.61, Revision 1 are used. These values are presented in Table 3.7.16. For the soil and rock materials, the damping ratio is obtained based on straincompatible soil properties generated for each soil profile. Soil material damping ratios are shown on Table 3.7.115 through Table 3.7.119 for each soil type considered. Soil damping ratio is limited to 15 percent.

RAI 03.07.02-11 The implementation of these damping values in the dynamic analyses of the NuScale RXB and CRB does not follow guidance from DSRS Section 3.7.2.II.13. Instead, damping procedures that are more suitable with the type of analysis performed are followed. For transient analysis with ANSYS, Rayleigh material damping is used. For soil-structure interaction analysis with SASSI2010, hysteretic material damping is used. Both Rayleigh and hysteretic damping provide responses equivalent to the composite modal damping approach. Only major components, such as the NPM and the RBC, are included in the dynamic models. For other systems and components, their mass is applied to the model and ISRS are calculated at the corresponding damping level in Table 3.7.1-6.

3.7.2.16 Site Specific Seismic Analysis RAI 03.08.04-23S1 Site-specific seismic analysis is performed by the COL applicant to confirm that the site-independent Seismic Category I structures may be constructed without modification, or to identify where modifications are necessary. This comparison is performed in Section 3.8.4.8. The site specific analysis is performed using the site specific SSE developed in Section 3.7.1.1.3 (COL Item 3.7-1) and the site specific soil profile developed in Section 3.7.1.3.3 (COL Item 3.7-3). Appendix 3B critical sections include RXB and CRB exterior walls that are subject to earth pressures. Therefore, by comparing seismic demand in these walls per COL Item 3.7-5, site-specific versus lateral certified standard soil pressures are also compared.

RAI 03.07.02-12, RAI 03.08.04-23S1, RAI 03.08.04-23S2 COL Item 3.7-5: A COL applicant that references the NuScale Power Plant design certification will perform a soil-structure interaction analysis of the Reactor Building and the Control Building using the NuScale SASSI2010 models for those structures. The COL applicant will confirm that the site-specific seismic demands of the standard design for critical structures, systems, and components in Appendix 3B are bounded by Tier 2 3.7-145 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design the corresponding design certified seismic demands and, if not, the standard design for critical structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. Seismic demands investigated shall include forces, moments, deformations, in-structure response spectra, and seismic stability of the structures.

RAI 03.07.02-12 COL Item 3.7-6: A COL applicant that references the NuScale Power Plant design certification will perform a structure-soil-structure interaction analysis that includes the Reactor Building, the Control Building, the Radioactive Waste Building and both Turbine Generator Buildings. The COL applicant will confirm that the site-specific seismic demands of the standard design structures, systems, and components are bounded by the corresponding design certified seismic demands and, if not, the standard design structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands.

3.7.2.17 References 3.7.2-1 SAP2000 Advanced Version 17.1.1, 2015, Computers and Structures, Inc.,

Walnut Creek, California.

3.7.2-2 SASSI2010 Version 1.0, May 2012, Berkeley, California.

3.7.2-3 ANSYS Computer Program, Release 16.0, January 2015. ANSYS Incorporated, Canonsburg, Pennsylvania.

3.7.2-4 American Society of Mechanical Engineers, ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), 2004.

Tier 2 3.7-146 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design the west of the bay. With the removal of the module for the 7 NPM case. The bending moments increased by 30 to 40 percent. This increase is attributed to the larger water volume. The Bay 6 pool wall was essentially unaffected. Bay 6 contains a module in the 7 NPM case.

3.7.2.9.1.5 Conclusion of the Study RAI 03.07.02-8 The 7 NPM case did not produce a tangible change in the reaction of the building as a whole (Section 3.7.2.9.1.1 and Section 3.7.2.9.1.2). The 6 NPM case, which would cause a slightly more asymmetric load, is expected to produce similar results. The mass of the overall structure is relatively unaffected by the mass difference between a NPM and the water. Therefore the quantity of modules installed in the building is expected to have no effect on the building.

Similarly, the absence of modules did not significantly affect the forces that are transmitted to an installed NPM (Section 3.7.2.9.1.3). Therefore removing individual modules for refueling does not impact the installed and operating modules.

The walls of bays without an installed module do see an increase in the forces, principally in bending moment. These increases are on the order of 40 percent.

However, the wing walls are all designed the same. As such, they are designed for the highest loaded wall, which is the west wing wall. The increases seen in the west wing wall when an NPM is not present in Bay 1 do not exceed the capacity of the wall. In addition, the increase is less significant because there is no module supported by the wall.

The pool wall in an empty bay also sees an increase of about 40 percent. Again, the highest forces occur at the west end of the pool. The forces at the pool wall in Bay 1 when it is empty are similar to those in the reactor pool area. Since the entire pool wall is a consistent design, these forces are also acceptable.

RAI 03.07.02-8 The difference in results between operation with twelve NPMs and operation with fewer NPMs in place is small and within the capacity of the building design. Site-specific configurations, outside of the scope of the presented 12 NPM and 7 NPM cases, require additional analysis to be performed by the COL applicant.

RAI 03.07.02-8 COL Item 3.7-10: A COL applicant that references the NuScale Power Plant design certification will perform a site-specific configuration analysis that includes the Reactor Building with applicable configuration layout of the desired NuScale Power Modules. The COL applicant will confirm the following are bounded by the corresponding design certified seismic demands:

RAI 03.07.02-8, RAI 03.08.04-23S2 Tier 2 3.7-147 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design

1) The in-structure response spectra of the standard design at the foundation and roof. See FSAR Figure 3.7.2-107 and Figure 3.7.2-108 for foundation in-structure response spectra and Figure 3.7.2-113 for roof in-structure response spectra.

RAI 03.07.02-8, RAI 03.08.04-23S2

2) The maximum forces in the NuScale Power Module lug restraints and skirts.

RAI 03.08.04-23S1, RAI 03.08.04-23S2

3) The site-specific in-structure response spectra for the NuScale Power Module at the skirt support will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-156 and Figure 3.7.2-157. The site-specific in-structure response spectra for the NuScale Power Module at the lug restraints will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-158 through Figure 3.7.2-163.

RAI 03.07.02-8, RAI 03.08.04-23S2

4) The maximum forces and moments in the east and west wing walls and pool walls. See FSAR Table 3.7.2-32.

RAI 03.08.04-23S1, RAI 03.08.04-23S2

5) The site-specific in-structure response spectra for the fuel storage racks will be shown to be bounded by the in-structure response spectra in Figure 3-6 through Figure 3-14 of TR-0816-49833.

RAI 03.08.04-23S1, RAI 03.08.04-23S2

6) The site-specific in-structure response spectra shown immediately below will be shown to be bounded by their corresponding certified in-structure response spectra:

RAI 03.08.04-23S1, RAI 03.08.04-23S2

  • Reactor Building north exterior wall at EL 75-0: bounded by in-structure response spectra in Figure 3.7.2-110 RAI 03.08.04-23S1, RAI 03.08.04-23S2
  • Reactor Building west exterior wall at EL 126-0: bounded by in-structure response spectra in Figure 3.7.2-112 RAI 03.08.04-23S1, RAI 03.08.04-23S2
  • Reactor Building crane wheels at EL 145-6: bounded by in-structure response spectra in Figure 3.7.2-114 RAI 03.08.04-23S1, RAI 03.08.04-23S2
  • Control Building east wall at EL 76-6: bounded by in-structure response spectra in Figure 3.7.2-119 RAI 03.08.04-23S1, RAI 03.08.04-23S2
  • Control Building south wall at EL 120-0: bounded by in-structure response spectra in Figure 3.7.2-121 RAI 03.07.02-8 If not, the standard design will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands.

Tier 2 3.7-148 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Table 3.7.2-53: Floor Elevation and Nodes for Floor ISRS Generation Floor No. TOC Note Standalone Triple Model Coordinates (inch)

Elevation RXB Node Node X Y Z 1 EL 24'-0" Top of 3996 3652 0 873 120 Basemat 4741 4325 1872 873 120 5642 5142 4092 873 120 2 EL. 25-0 Pool Floor 6041 5525 2019.5 305.5 132 (NPM Base) 6093 5577 2314.5 305.5 132 6145 5629 2609.5 305.5 132 6197 5681 2904.5 305.5 132 6249 5733 3199.5 305.5 132 6301 5785 3509.5 305.5 132 6065 5549 2167 177 132 6013 5497 1872 177 132 6069 5553 2167 453 132 6017 5501 1872 453 132 6325 5809 3672 177 132 6273 5757 3347 177 132 6329 5813 3672 453 132 6277 5761 3347 453 132 6317 5801 3672 -453 132 6265 5749 3347 -453 132 6321 5808 3672 -177 132 6269 5753 3347 -177 132 6057 5541 2167 -453 132 6005 5489 1872 -453 132 6061 5545 2167 -177 132 6009 5493 1872 -177 132 3 EL. 50-0 10974 9955 0 873 420 11050 10022 216 0 420 11054 10026 216 279 420 11234 10185 824 705 420 11542 10451 1872 453 420 11675 10566 2314.6 621 420 11995 10844 3347 621 420 12174 11002 3924 88.5 420 12178 11006 3924 360 420 12242 11067 4092 873 420 4 EL. 75-0 16925 14941 0 -873 720 16947 14963 0 873 720 17207 15193 824 705 720 17630 15556 2314.5 621 720 17942 15826 3347 621 720 18031 15903 3672 453 720 18123 15986 3924 360 720 Tier 2 3.7-255 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Table 3.7.2-53: Floor Elevation and Nodes for Floor ISRS Generation (Continued)

Floor No. TOC Note Standalone Triple Model Coordinates (inch)

Elevation RXB Node Node X Y Z 5 EL. 100-0 Grade Floor 22810 19886 0 -837 1020 22821 19897 0 0 1020 22832 19908 0 837 1020 22905 19972 216 -228 1020 23092 20138 824 705 1020 23517 20503 2314.5 621 1020 23829 20773 3347 621 1020 24008 20931 3924 88.5 1020 24012 20935 3924 360 1020 23386 20390 1872 453 1020 23915 20847 3672 177 1020 23919 20851 3672 453 1020 6 EL. 126-0 25487 22328 0 -873 1320 25509 22350 0 873 1320 25625 22466 824 705 1320 25826 22667 1872 453 1320 25831 22672 1872 873 1320 25952 22793 2314.5 621 1320 26258 23099 3347 621 1320 26345 23186 3672 453 1320 26419 23260 3924 88.5 1320 26423 23264 3924 360 1320 26471 23312 4092 873 1320 Roof EL. 181-0 Top of Roof 29953 26794 0 -537 1980 29960 26801 0 0 1980 29967 26808 0 537 1980 30110 26951 824 0 1980 30350 27191 2019.5 0 1980 30357 27198 2019.5 537 1980 30515 27356 2830.75 0 1980 30748 27589 4092 -537 1980 30755 27596 4092 0 1980 30762 27603 4092 537 1980 Tier 2 3.7-256 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Table 3.7.2-54: SASSI CNV Skirt Nodes NPM1 NPM6 6027 6287 6028 6288 6029 6289 6039 6299 6042 6302 6053 6307 6054 6308 6055 6309 Tier 2 3.7-257 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Table 3.7.2-55: SASSI CNV Lug Nodes NPM1 NPM6 West Lug 6477 31081 North Lug 6483 31087 East Lug 6486 31090 Tier 2 3.7-258 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-142: Floor ISRS Locations at TOC EL 24'-0" RAI 03.08.04-23S2 Tier 2 3.7-392 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-143: Floor Locations at TOC EL 25-0 RAI 03.08.04-23S2 Tier 2 3.7-393 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-144: Floor ISRS Locations at TOC EL 50'-0" RAI 03.08.04-23S2 Tier 2 3.7-394 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-145: Floor ISRS Locations at TOC EL 75' - 0" RAI 03.08.04-23S2 Tier 2 3.7-395 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-146: Floor ISRS Locations at TOC EL 100'-0" RAI 03.08.04-23S2 Tier 2 3.7-396 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-147: Floor ISRS Locations at TOC EL 126'-0" RAI 03.08.04-23S2 Tier 2 3.7-397 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design Figure 3.7.2-148: Roof ISRS Locations at TOC EL 181'-0" RAI 03.08.04-23S2 Tier 2 3.7-398 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-156: Enveloping ISRS of Cases 1, 2, and 3 at the CNV Skirt of NPM1 Tier 2 3.7-414 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-157: Enveloping ISRS of Cases 1, 2, and 3 at the CNV Skirt of NPM6 Tier 2 3.7-415 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-158: Enveloping ISRS of Cases 1, 2, and 3 at the East Lug of NPM1 Tier 2 3.7-416 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-159: Enveloping ISRS of Cases 1, 2, and 3 at the North Lug of NPM1 Tier 2 3.7-417 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-160: Enveloping ISRS of Cases 1, 2, and 3 at the West Lug of NPM1 Tier 2 3.7-418 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-161: Enveloping ISRS of Cases 1, 2, and 3 at the East Lug of NPM6 Tier 2 3.7-419 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-162: Enveloping ISRS of Cases 1, 2, and 3 at the North Lug of NPM6 Tier 2 3.7-420 Draft Revision 2

NuScale Final Safety Analysis Report Seismic Design RAI 03.08.04-23S2 Figure 3.7.2-163: Enveloping ISRS of Cases 1, 2, and 3 at the West Lug of NPM6 Tier 2 3.7-421 Draft Revision 2

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