ML24354A160

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LLC, Response to NRC Request for Additional Information No. 027 (RAI-10189 R1) on the NuScale Standard Design Approval Application
ML24354A160
Person / Time
Site: 05200050
Issue date: 12/19/2024
From: Shaver M
NuScale
To:
Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML24354A159 List:
References
RAIO-177459
Download: ML24354A160 (1)
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RAIO-177459 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com December 19, 2024 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information No. 027 (RAI-10189 R1) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter to NuScale, Request for Additional Information No. 027 (RAI-10189 R1), dated July 16, 2024 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The enclosure to this letter contains the NuScale response to the following RAI question from NRC RAI-10189 R1:

5.4.1.6.1-1 is the proprietary version of the NuScale Response to NRC RAI No. 027 (RAI-10189 R1, Question 5.4.1.6.1-1). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.

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

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

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 19, 2024.

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

RAIO-177459 Page 2 of 2 12/19/2024 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Stacy Joseph, Senior Project Manager, NRC

NuScale Response to NRC Request for Additional Information RAI-10189 R1, Question 5.4.1.6.1-1, Proprietary Version : NuScale Response to NRC Request for Additional Information RAI-10189 R1, Question 5.4.1.6.1-1, Nonproprietary Version : Affidavit of Mark W. Shaver, AF-177460

RAIO-177459 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10189 R1, Question 5.4.1.6.1-1, Proprietary Version

RAIO-177459 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10189 R1, Question 5.4.1.6.1-1, Nonproprietary Version

Response to Request for Additional Information Docket: 052000050 RAI No.: 10189 Date of RAI Issue: 07/16/2024 NRC Question No.: 5.4.1.6.1-1 Regulatory Basis

  • 10 CFR 50.36, Technical Specifications, requires, in part, limiting conditions for operation and administrative controls for maintaining integrity of a fission product barrier.
  • 10 CFR 50.55a, Codes and Standards, requires that throughout the service life of a PWR, Class 1 components must meet the requirements, except design and access provisions and PSI requirements, in Section XI, to the extent practical.
  • 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, requires that licensees monitor the performance or condition of SSCs against goals to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions.

Issue Section 16.1.1, Introduction to Technical Specifications, of the Standard Design Approval Application (SDAA) states, These revised GTS [Generic Technical Specifications] were developed consistent with the Improved Standard Technical Specification (ISTS) format and content typified in NUREG-1431 Standard Technical Specifications - Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases, Revision 5, dated September 2021 (ML21259A155 and ML21259A159, respectively), and NUREG-1432, [Standard Technical NuScale Nonproprietary NuScale Nonproprietary

Specifications - Combustion Engineering Plants, Volume 1, Specifications, and Volume 2, Bases,] Revision 5, dated September 2021 (ML21258A421 and ML21258A424, respectively).

The ISTS include a plugging criterion in Section 5.5.8, Steam Generator (SG) Program, as part of maintaining structural and leakage integrity of the SG tubes in accordance with the ASME Code (10 CFR 50.55a). The criterion is defined in terms of a percentage of the nominal tube wall thickness and requires that tubes found by inservice inspection to contain flaws with a depth equal to or exceeding the plugging criterion must be plugged (or repaired if a licensee has an NRC-approved alternate repair criteria).

For the Design Certification Application (DCA), NuScale proposed a plugging criterion value of 40 percent of the nominal tube wall thickness, based on design factors considered to be bounding, and determined using a finite element analysis (FEA) in accordance with Regulatory Guide (RG) 1.121, Bases for Plugging Degraded PWR Steam Generator Tubes (ML003739366). In addition, the 40 percent plugging criterion was bracketed in the DCA GTS to ensure plant-specific factors would be addressed by Combined License (COL) applicants.

For the SDAA, NuScale proposed a value of 40 percent of nominal tube wall thickness as a tube plugging criterion in the GTS and Sections 5.4.1.4 and 5.4.1.6.1 of the Final Safety Analysis Report (FSAR); however this value is not bracketed in the GTS. During the SDA audit, NuScale stated that the 40 percent plugging criterion is intended to apply to all plants referencing the SDAA, providing only general qualitative statements for considering this value, lacking adequate justification. NuScale did not address the staffs questions about whether the FEA used to justify the value of 40 percent in the NuScale DCA had been updated for SDAA design changes, i.e.,

whether any other analyses have been performed to support the value of 40 percent, or how the updated analysis or any other analyses considered applicable loads, degradation mechanisms, degradation growth, nondestructive examination uncertainty, and tube ovality.

NuScale stated that the FEA performed for the US600 in the DCA has not been revised to reflect the US460 design in the SDAA. NuScale also stated that the geometry and tube material are the same as for the US600 but acknowledged that the different operating conditions for the US460 may affect the plugging criterion. However, because the SDAA proposes a different design for SG tube supports, the staff disagreed with NuScale that the geometry to be considered in the plugging criterion determination has not changed. In response, NuScale proposed using a bracketed value of [40%] in the GTS and deleting the value from FSAR Sections 5.4.1.4 and 5.4.1.6.1.

NuScale Nonproprietary NuScale Nonproprietary

The staff considers NuScales proposal to be inadequate to make a finding on the plugging criterion. The staff acknowledges the benefit of bracketing the value, designating it as COL information that must be finalized by each COL applicant referencing the US460 SDAA.

Therefore, consistent with DSRS Chapter 16, the staff needs additional information as specified below to make a finding that the value proposed in the SDAA reasonably agrees with the expected operational capability of the plant.

Information Requested

1. Provide an analysis, including any available experimental information, that supports the US460 SG tube plugging criterion for staff audit. The analysis should demonstrate how the regulatory positions in RG 1.121 are met.
2. Provide corresponding markups to show how the SG tube plugging criterion will be incorporated into the FSAR, including a summary of the analysis results, and GTS.

Feedback received on November 7, 2024:

Information requested in docketed correspondence (e.g., RAI response)

a. Identify the amount of ovality assumed in the tube plugging limit analysis and how it compares to the ovality limits specified for the US460 steam generator design.
b. A programmatic basis was provided for not considering cracking in the evaluation of the tube plugging criterion (i.e., plug on detection). Provide the technical basis for not evaluating cracking.

Actions the NRC understands NuScale is taking

a. FSAR Section 5.4.1.4 will be revised for consistency with Technical Specifications wording on the tube plugging criterion.
b. The description of the proposed new structural integrity performance criterion in Section 5.5.4.b.1 of the Generic Technical Specifications will be clarified.
c. Consideration will be given to revising FSAR Table 1.9-2 or FSAR Section 5.4.1 to clarify the exceptions being taken to RG 1.121.

NuScale Response:

To determine an appropriate steam generator tube plugging criterion for the US460 design, NuScale performed a finite element analysis specific to the US460 design and provided the NuScale Nonproprietary NuScale Nonproprietary

analysis to the NRC staff. The analysis uses the Nuclear Energy Institute (NEI) Steam Generator Program Guidelines (NEI 97-06, Revision 3), which references the Electric Power Research Institute (EPRI) Steam Generator Integrity Assessment Guidelines to support the intent of Regulatory Guide (RG) 1.121. The NEI and EPRI guidelines assume that the primary coolant pressure is inside the steam generator tubes, so the analysis follows these guidelines with some adjustments in order to be applicable to and reasonable for the US460 design. The US460 finite element analysis meets the intent of RG 1.121; the analysis shows reasonable assurance of steam generator tube integrity and is consistent with the US600 design calculation.

To apply the NEI and EPRI guidelines to the US460 design, NuScale partially conforms to RG 1.121; NEI 97-06, Revision 3; and EPRI Steam Generator Integrity Assessment Guidelines acceptance criteria. The tube structural integrity performance criterion (SIPC) historically uses a safety factor of three on the normal operating pressure differential (3 NOPD), which is not appropriate for the US460 design. The US460 analysis uses an alternate safety factor.

Justification for the alternate NOPD safety factor, as well as details of the US460 failure criteria and design considerations, are addressed below. The US460 tube plugging criterion analysis conforms to the remaining loading condition and safety factor guidance in the SIPC.

Failure Criteria The limiting failure mechanism for the NuScale designs during normal operating conditions is buckling under compression because primary coolant pressure is outside the steam generator (SG) tubes. The NEI and EPRI guidelines direct analysis of burst and collapse under tension and bending. Buckling occurs at a much lower strain than the strain at ultimate strength considered for tube burst.

During a tube failure by burst, the fission product barrier is compromised. During a tube failure by buckling (a form of plastic collapse), the failed tube may not rupture and thus may still maintain the fission product barrier.

Consideration of alternate failure criteria for the US460 design is appropriate because the primary coolant pressure is outside the tubes, and the associated mechanism of tube failure by buckling is not addressed by RG 1.121. Consideration of the failure mechanism most appropriate for the design meets the intent of RG 1.121.

NuScale Nonproprietary NuScale Nonproprietary

Alternative Loading NuScale uses a proposed alternative to the tube structural integrity performance criterion loading condition of 3 NOPD. The basis for the 3 NOPD loading condition is not applicable to the US460 design because RG 1.121, NEI, and EPRI recommendations consider SG tube integrity due to failure mechanisms resultant from the internal pressure of the tube being greater than the external pressure. The US460 design is not susceptible to SG tube burst due to normal operating differential pressure because the external pressure on the tube is greater than the internal pressure within the tube.

Regulatory Position C.3.(a)(2) of RG 1.121 states:

The margin between the maximum internal pressure to be contained by the tubes during normal plant conditions and the pressure that would be required to burst the tubes should remain consistent with the margin incorporated in the design rules of Section III of the ASME Code.

While tube burst due to internal pressure is not a credible tube failure mechanism for the US460 design, RG 1.121 considered margin developed in the American Society of Mechanical Engineers (ASME) Code to be sufficient to meet the intent of the guidance. NuScale proposes a safety factor of 2.0 against buckling or collapse at the NOPD. This safety factor is consistent with design factors found in the ASME Code and therefore aligns with the RG 1.121 perspective on ASME Code margin.

NuScale uses ASME Code Case N-759-2 to determine the minimum SG tube wall thickness, which provides alternative rules for determining allowable external pressure and compressive stresses in various components in lieu of the rules of ASME Section III, Division 1, NB-3133.

Section 2.2 of ASME Code Case N-759-2 uses stress reduction factors of 1.67 to 2.0. ASME Section VIII, Division 2, Part 4, 4.4, Design of Shells Under External Pressure and Allowable Compressive Stresses is very similar to Code Case N-759-2 and uses design factors of 1.67 to 2.0 in Section 4.4.2, Design Factors. ASME Section VIII, Division 2, Part 5, 5.4.3, Protection Against Collapse from Buckling: Buckling Analysis - Method B, requires a multiplier of 1.67 to be applied to the pressure-plus-deadweight load combination when elastic-plastic analysis is performed to demonstrate buckling is prevented. Finally, ASME Section III, Mandatory Appendix XIII, XIII-3200 provides an option for plastic analysis as an alternative to meeting membrane and membrane plus bending stress intensity limits of XIII-3130. XIII-3200 applies a factor for the collapse load of 2/3 for Service Level A and Service Level B, which corresponds to a design factor of 1.5.

NuScale Nonproprietary NuScale Nonproprietary

While multiple portions of the ASME Code incorporate design factors against collapse and buckling of between 1.5 and 2, as described above, the external pressure design rules of NB-3133 use a design factor of 3. The design factor of 3 is described in ASME Section II, Part D, Mandatory Appendix 3. Paragraph 3-600(a):

For cylindrical shells, under external pressures, the allowable stress is the least of (1) 33% of the assumed critical buckling stress... (2) 33% of the specified minimum yield strength and yield strength at temperature...

This supports a design factor of 3 for buckling when sizing the component; however, this design factor for buckling is high due to the NB-3133 hand calculation approach using charts and tables that originated in the 1940s. NuScale uses Code Case N-759-2 instead of NB-3133 for determining SG tube wall thickness under design conditions and a finite element analysis to calculate the external pressure capacity of a degraded tube for comparison to the tube plugging limit. Therefore, it is appropriate to consider design factors in the ASME Code consistent with the types of analysis performed. Based on the review of the various design factors for buckling or collapse described above, a limit of 2.0 NOPD is appropriate for the US460 design as noted in the attached Technical Specifications markups.

The buckling pressure determined in the US460 finite element analysis exceeds the 2.0 NOPD criterion ((2(a),(c) The analysis conclusion supports a 40 percent through-wall thickness SG tube plugging criterion. The attached markup brackets this SG tube plugging criterion and describes the alternate loading condition as at least 2.0 times the NOPD for buckling or collapse in the Technical Specifications. The analysis shows that the US460 SG design has an external pressure capacity greater than 2.0 times NOPD. The partial conformance with RG 1.121 is outlined in Final Safety Analysis Report Table 1.9-2, and Final Safety Analysis Report (FSAR) Chapter 5 markups consistently describe the SG tube plugging criterion for the US460 design. This response describes the exceptions to RG 1.121 and to the NEI and EPRI guidance. The FSAR does not require updating beyond identifying partial conformance with RG 1.121 in Table 1.9-2, consistent with treatment of other documents. NuScale Nonproprietary NuScale Nonproprietary

Tube Thickness The SG tube thickness for the US460 design is thicker than existing designs, as shown in Table 5.4-2 of the FSAR. One of the concerns of RG 1.121 is the SG tube thickness. Use of a thicker SG tube lessens the likelihood of SG tube failure and meets the intent of RG 1.121. Ovality The maximum SG tube ovality permitted by the NuScale design specification is (( 

}}2(a),(c). The analysis assumes the maximum ovality of

(( 

}}2(a),(c) at all locations.

Cracking The EPRI Steam Generator Integrity Assessment Guidelines state that a plug-on-detection repair scenario... is typically applied to crack-like degradation. The implementation of the Steam Generator Program, which ensures SG tube integrity as noted in Technical Specifications 5.5.4, will be developed and implemented to address Combined Operating License (COL) Item 5.4-1. The COL Item 5.4-1 commits to basing the Steam Generator Program on NEI 97-06 and the EPRI steam generator guideline revisions available at the time of the COL application. If SG tubes are plugged on detection of a crack, the technical basis for the SG tube plugging criterion does not need to consider cracks. Depending on the results of the degradation assessment (Section 5.4.1.6), a COL applicant could perform a crack stability analysis and a structural integrity limit analysis to support the license-specific SG tube plugging criterion as part of the Steam Generator Program. This will substantiate the SG tube plugging limit for the COL applicant. For the US600 design, NuScale concluded that cracks are not considered unless coupled with seismic loading due to external pressure on the tubes. Crack opening and propagation can only occur in tubes subjected to tensile stresses. External pressure places SG tubes in a compressive field; therefore, a crack will not open unless additional loading develops a net tensile stress at the crack location. For the US600 design, NuScale assessed (( 

}}2(a),(c) and found that they were stable at

(( 

}}2(a),(c)

NuScale Nonproprietary NuScale Nonproprietary

Impact on US460 SDAA: FSAR Section 1.9, Section 5.4, and Technical Specifications Section 5.5 have been revised as described in the response above and as shown in the markups provided in this response. Technical report TR-101310 is updated to address the change from Standard Technical Specifications as shown in the markups provided with this response. NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Conformance with Regulatory Criteria NuScale US460 SDAA 1.9-3 Draft Revision 2 Audit Question A-3.5.1.3-2, Audit Question A-3.7.3-3, Audit Question A-3.11.2.3-1, Audit Question A-5.2.3.4.2-1, Audit Question A-6.1.1-2, Audit Question A-6.1.1-8, Audit Question A-6.2.5-1, Audit Question A-8.1-4, Audit Question DWO-SC-26, Audit Question EDAS Deep Dive Action Item 1, Audit Question EDAS Deep Dive Action Item 3, Audit Question EDAS Deep Dive Action Item 4, Audit Question EDAS Deep Dive Action Item 5, Audit Question EDAS Deep Dive Action Item 6, Audit Question EDAS Deep Dive Action Item 9, Audit Question EDAS Deep Dive Action Item 11, Audit Question EDAS Deep Dive Action Item 14 RAI 5.4.1.6.1-1, RAI 19.2-1, RAI 19.2-3, RAI 19.2-4 Table 1.9-2: Conformance with Regulatory Guides RG Title Rev. Conformance Status Comments Section 1.6 Safety Guide 6 - Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Sys-tems 0 Not Applicable The onsite electrical AC power systems do not contain Class 1E distribution systems. Not Applicable 1.7 Control of Combustible Gas Con-centrations in Containment 3 Partially Conforms The design complies with the intent of RG 1.7 regulatory positions that address atmosphere mixing, hydrogen gas production, and containment structural integrity. However, the design deviates from the positions on hydrogen and oxygen monitors. The design includes a passive autocatalytic recombiner (PAR) that is sized to limit oxygen concentrations to a level that does not sup-port combustion (i.e., less than four percent), this results inmaintaining an inert containment atmosphere. The design and quality standards applied to the PAR are commensurate with its safety-related, non-risk-signifi-cant function in the NuScale design, rather than the non-safety-related, risk-significant function underlying regulatory position C.1. The NuScale design does not include combustible gas monitoringsupports an exemp tion to 10 CFR 50.44(c)(4). 6.2.5 1.8 Qualification and Training of Per-sonnel for Nuclear Power Plants 4 Not Applicable This guidance governs site-specific programmatic and operational activities that are the responsibility of the applicant or licensee. Not Applicable 1.9 Application and Testing of Safety-Related Diesel Genera-tors in Nuclear Power Plants 4 Not Applicable The NuScale design does not require or include safety-related emergency diesel generators. Not Applicable 1.11 Instrument Lines Penetrating the Primary Reactor Containment 1 Not Applicable No instrument lines penetrate the NuScale Power Mod-ule (NPM) containment. Not Applicable

NuScale Final Safety Analysis Report Conformance with Regulatory Criteria NuScale US460 SDAA 1.9-17 Draft Revision 2 1.113 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appen-dix I 1 Not Applicable This guidance governs analysis of the aquatic disper-sion of radioactive liquid effluents from component fail-ures, in accordance with BTP 11-6. Because the NuScale facility provides an approved design mitigative feature (metal-lined concrete dike around the pool surge control subsystem storage tank), such an analysis is not required. Not Applicable 1.114 Guidance to Operators at the Controls and to Senior Operators in the Control Room of a Nuclear Power Unit 3 Partially Conforms This guidance is applicable except for site-specific guid-ance that is the responsibility of the applicant or licensee. Consistent with the discussion in RG 1.114, Section B.1, the ability of the applicant to meet this guid-ance is facilitated by the control room design and layout (including the designated surveillance area described in Position C.1.3). Portions of this guidance that implement operator staffing requirements of 10 CFR 50.54(m)(2)(i) and (iii) (e.g., Position C.1.5) are not applicable to appli-cants. 18.5 1.115 Protection Against Turbine Mis-siles 2 Conforms None. 3.5 1.117 Protection Against Extreme Wind Events and Missiles for Nuclear Power Plants 2 Conforms Confirmation of site-specific design-basis automobile missile parameters and site proximity missilesthat nearby structures exposed to extreme wind loads will not adversely affect the RXBor the Seismic Category I portion of the Control Building is the responsibility of the applicant or licensee. This guidance is not applicable to the CRB, however, the Seismic Category I portions of the CRB conform to relevant guidance of RG 1.117. 3.5 9.1.2 1.118 Periodic Testing of Electric Power and Protection Systems 3 Partially Conforms This guidance is applicable except for site-specific guid-ance that is the responsibility of the applicant or licensee. 7.2 14.2 1.121 Bases for Plugging Degraded PWR Steam Generator Tubes 0 Partially Conforms None.This guidance assumes primary coolant pressure inside the steam generator tubes, while the NuScale design has primary coolant pressure outside the tubes. The design complies with the intent of the guidance but uses an alternate loading condition. 5.4 Table 1.9-2: Conformance with Regulatory Guides (Continued) RG Title Rev. Conformance Status Comments Section

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-10 Draft Revision 2 current test meets Electric Power Research Institute (EPRI) 1013706 (Reference 5.4-2). Preservice examinations performed in accordance with the ASME BPVC, Section III, Subsubarticle NB-5280 and Section XI, Subarticle IWB-2200 (Reference 5.4-5) use examination methods of ASME BPVC Section V, except as modified by Section III, Paragraph NB-5111. These preservice examinations include essentially 100 percent of the pressure boundary welds. Audit Question A-5.4.1.4-1, Audit Question A-5.4.1.6.1-1, Audit Question DWO-SC-26 RAI 5.4.1.6.1-1 A preservice volumetric, full-length preservice inspection of essentially 100 percent of the tubing in each SG is performed. The length of the tube extends from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet welds are not part of the tube. The preservice inspection is performed after tube installation and shop or field primary-side hydrostatic testing and before initial power operation to provide a definitive baseline record, against which future ISI can be compared. Technical Specifications Section 5, Administrative Controls, defines the tube plugging criterion as the maximum allowable flaw in the tube wall. Tubes with flaws that are equal to or exceed 40 percent of the nominal tube wall thicknessthe tube plugging criterion are plugged. Tubes with flaws that could potentially compromise tube integrity before the performance of the first ISI, and tubes with indications that could affect future inspectability of the tube, are also plugged. The volumetric technique used for the preservice examination is capable of detecting the types of preservice flaws that may be present in the tubes and permits comparisons to the results of the ISI expected to be performed to satisfy the SG tube inspection requirements in accordance with the plant technical specifications. Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 As discussed above, the operational inservice testing and inspection programs described in Section 5.2.4, RCPB ISI and Testing, and the SG program described in Section 5.4.1.6, Steam Generator Program, provide testing and inspection requirements following initial plant startup. The SG inlet flow restrictors are examined by VT-3 in accordance with IWA-2213 when removed for SG tube examinations. Inservice inspection and testing of the SGS steam and feedwater piping is described in Section 6.6. 5.4.1.5 Steam Generator Materials Selection and fabrication of pressure boundary materials used in the SGs and associated components are in accordance with the requirements of ASME BPVC Section III and Section II as described in Section 5.2.3, RCPB Materials, and the materials used in the fabrication of the SGs are in Table 5.2-3. The RCPB materials used in the SGS are Quality Group A and their design, fabrication, construction, tests, and inspections conform to Class 1 in accordance with the ASME BPVC and the applicable conditions promulgated in 10 CFR 50.55a(b). The SGS materials forming the RCPB, including weld materials, conform to fabrication, construction, and testing requirements of ASME

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-13 Draft Revision 2 degradation allowance (additional tube wall thickness above minimum required for design) as discussed in Section 5.4.1.2, System Design. The NPM reactor coolant flowrates are also lower than the flowrates across the SG tubes in PWR recirculating SGs as discussed in Section 5.1, RCS and Connecting Systems. This low flow rate reduces the flow energy available to cause FIV wear degradation of SG tubes. Based on the additional tube wall margin and the additional margin against FIV turbulent buffeting wear (the most likely SG tube degradation mechanism), application of the existing PWR SG Program requirements to the design is appropriate. For SGs in the PWR fleet with SB-163 UNS N06690 SG tubing, the only observed degradation has been wear as a result of FIV (tube-to-tube or tube-to-support plate) or wear due to foreign objects. With respect to the risk of introduction of foreign objects, the NPM is at no greater risk than existing designs; therefore, the design does not warrant deviations from existing SG program guidelines. From the standpoint of SG tube design, the two significant differences between the SG design and current large PWR designs is the helical shape of the SG tubing and the SG tube support structure. The helical shape of the SG tubing itself does not represent risk of degradation based on the minimum bend radius of the helical tubing being within the historical experience base of PWR SG designs. Prototypic testing of the SG tube supports validates acceptable performance (including wear) of the SG tube support design. Implementation of a typical SG program is appropriate based on evaluation of the design of the SG tube supports. 5.4.1.6.1 Degradation Assessment Audit Question A-5.4.1.6.1-1, Audit Question DWO-SC-26 RAI 5.4.1.6.1-1 A degradation assessment of the NPM SG identifies several potential degradation mechanisms. Wear is the most likely degradation mechanism, and there is the potential for several secondary side corrosion mechanisms, including under deposit pitting and intergranular attack based on the once-through design with secondary boiling occurring inside the tubes. The estimated growth rates for these potential defects is sufficiently low that the SG tube plugging criterion for the SG is a 40 percent through wall defect. Operational SG tube integrity is ensured by implementing tube plugging criteria, implementing elements of the SG program, and implementing the SG inspections. Audit Question A-3.9.2-26-F, Audit Question A-3.9.2-28, Audit Question DWO-SC-23, Audit Question DWO-SC-24, Audit Question DWO-SC-36, Audit Question DWO-SC-37 A 100 percent SG tube inspection is completed during the first refueling outage following initial startup or SG replacement. After the first refueling outage, a 100 percent SG inspection is completed on a staggered basis over the next 72 effective full power months in order to evaluate ongoing SG tube degradation. COL Item 5.4-1: An applicant that references the NuScale Power Plant US460 standard design will develop and implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that

Programs and Manuals 5.5 NuScale US460 5.5-4 Draft Revision 2 5.5 Programs and Manuals 5.5.4 Steam Generator (SG) Program (continued)

a.

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion: All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, and cool down and all anticipated transients included in the design specification) and design basis accidents. Structural integrity is defined as no tube failure through gross structural deformation in burst, collapse, or buckling. This includes retaining a safety factor of greater than 2.03.0 against burstcollapse or buckling under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burstfailure applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapsefailure. In the assessment of tube integrity, those loads that do significantly affect burst or collapsefailure shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube failure, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 150 gallons per day.

Programs and Manuals 5.5 NuScale US460 5.5-5 Draft Revision 2 5.5 Programs and Manuals 5.5.4 Steam Generator (SG) Program (continued)

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.5, "RCS Operational LEAKAGE.

c.

Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding [40%] of the nominal tube wall thickness shall be plugged.

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following initial startup or SG replacement.

2.

After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 7296 effective full power months, which defines the inspection period.

3.

If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected unit SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

US460 Standard Design Approval Technical Specifications Development TR-101310-NP Draft Revision 1 © Copyright 2024 by NuScale Power, LLC 16 3.4 Chapter 4, Design Features Section 4.3 Fuel Storage The fuel storage design description is modified to reflect changes to the design and analyses. Key variables are bracketed to allow replacement with actual plant-specific values when design details are finalized by a future applicant that references the NuScale power plant US460 standard design. NuScale is monitoring industry efforts to relocate fuel storage detailed requirements to a COLR-like document and anticipates adopting this practice when the concept matures. 3.5 Chapter 5, Administrative Controls Section 5.2.2, Facility Staff This section is modified to reflect approved topical report TR-0420-69456, NuScale Control Room Staffing Plan, TR-0420-69456, Revision 1-A. RAI 5.4.1.6.1-1 Section 5.5.4, Steam Generator (SG) Program This section is modified to more accurately reflect the design of the steam generator with primary coolant pressure outside the tubes and updates the safety factor for primary-to-secondary pressure differential. Section 5.5.9, Containment Leakage Rate Testing Program The description of the Containment Leakage Rate Testing Program is revised to provide alternatives to adopt Option A or Option B of 10 CFR 50, Appendix J. Section 5.6.3, Core Operating Limits Report This section is modified to align with the safety analyses, referencing technical specification limits, and topical reports that describe the limits that will be included in the COLR. Section 5.6.4, Reactor Coolant System PRESSURE AND TEMPERATURE LIMITS REPORT Modified to align with the safety analyses and topical report that describes the limits that will be included in the Pressure and Temperature Limits Report.

RAIO-177459 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Affidavit of Mark W. Shaver, AF-177460

AF-177460 Page 1 of 2

NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the response by which NuScale develops its NuScale Power, LLC Response to NRC Request for Additional Information (RAI No. 10189 R1, Question 5.4.1.6.1-1) on the NuScale Standard Design Approval Application. NuScale has performed significant research and evaluation to develop a basis for this response and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScales competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScales intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed response to NRC Request for Additional Information RAI 10189 R1, Question 5.4.1.6.1-1. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document.

AF-177460 Page 2 of 2 (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScales technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on December 19, 2024. Mark W. Shaver}}

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