ML25031A444

LLC Submittal of NuScale LOCA Break Safety Case and Associated Chapter 19 Markups

ML25031A444
Person / Time
Site:	05200050
Issue date:	01/31/2025
From:	Griffith T NuScale
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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ML25031A443	List: ML25031A444
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LO-175581
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LO-175581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 31, 2025 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 Submittal of NuScale LOCA Break Safety Case and Associated Chapter 19 Markups During the third quarter 2024 SES meeting for the US460 SDAA, held on July 18, 2024, NuScale provided the NRC a position on the safety for CVCS and ECCS LOCA Break.

NuScale subsequently posted an updated revision of this safety case to the electronic reading room for the staffs review, on December 18, 2024. NuScale posted associated markups for Chapter 19 on December 31, 2024. NuScale Power, LLC (NuScale) hereby submits NuScale LOCA Break Safety Case and NuScale Safety Case Updated Markup (Chapter 19) for docketing in support of the Standard Design Approval Application review. contains the proprietary version of the report entitled, NuScale LOCA Break Safety Case. 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 4) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirement of 10 CFR 810. Enclosure 2 contains the nonproprietary version of the safety case.

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

If you have any questions, please contact Amanda Bode at (541) 452-7971 or at ABode@nuscalepower.com.

Sincerely, Thomas Griffith Manager, Licensing NuScale Power, LLC

LO-175581 Page 2 of 2 01/31/2025 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution:

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

NuScale LOCA Break Safety Case, Proprietary Version :

NuScale LOCA Break Safety Case, Nonproprietary Version :

NuScale Safety Case Updated Markup (Chapter 19), Nonproprietary :

Affidavit of Mark W. Shaver, AF-175582

LO-175581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale LOCA Break Safety Case, Proprietary Version

LO-175581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale LOCA Break Safety Case, Nonproprietary Version

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 1 of 8 NuScale LOCA Safety Case Executive Summary The US460 is designed to eliminate, to the extent practical, design-basis events, and to mitigate those that remain with simplified, passive safety systems. To this end, with respect to loss-of-coolant accidents (LOCAs), the design excludes traditional large-bore reactor coolant system (RCS) piping, limits reactor coolant pressure boundary (RCPB) piping and welds, applies robust design and inspection criteria, and uses simple, passive safety systems to mitigate accident sequences. Additionally, ((2(a),(c) for a postulated break at the emergency core cooling system (ECCS) valve flanges or due to a break between the containment vessel (CNV) and chemical and volume control system (CVCS) containment isolation valves (CIVs). What Can Go Wrong? Chemical and Volume Control System There are four lines that support CVCS functions and are part of the RCPB that penetrate the CNV; for each line there are three welds and a test fixture, but no pipe, between the CNV wall and the CIVs. The RCPB segments outside the CNV are robustly designed to preclude breach, but a hypothetical rupture of those welds or test fixture body would cause an unisolable blowdown of the RCS that is not within the design basis of the plant. Emergency Core Cooling System All RCPB pipe is located within the leak-tight CNV designed to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section III Class 1 rules (the same as reactor vessels), which serves to both contain fission products and provide an alternate core cooling path when the ECCS is actuated. The ECCS valves are flanged components bolted directly to the reactor pressure vessel (RPV). An instantaneous and catastrophic rupture of an ECCS valve flange could result in a larger flow path than the inadvertent opening of the ECCS valves analyzed in the design basis. How Likely Is It? Chemical and Volume Control System The RCPB components outside the CNV have numerous features intended to preclude rupture. NuScale Nonproprietary

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 2 of 8 The length is the minimum possible; pipe is excluded and the CIVs are welded directly to an isolation test fixture (necessary for Appendix J leak-rate testing), which is welded directly to a CNV safe-end. The welds and components are designed to satisfy the same stringent ASME code provisions as the RPV. The welds and components are further designed to Branch Technical Position (BTP) 3-4, requiring even more conservative stress and cumulative usage factors than the ASME code. Preliminary stress analyses completed for most of the welds show ((

}}2(a),(c)

Programmatic requirements, such as Inservice Inspection (ISI) and chemistry controls, prevent degradation of welds and components. The welds and components are rigorously inspected during service, to detect any possible degradation long before it could lead to sudden failure. Estimating the likelihood of an RCPB component rupture is not straightforward, due in large part to the infrequency of such major failures.1 Pipe failure data predominantly reflects leaks rather than major failures, and in the case of small lines includes instrument lines and steam generator tubes that are a poor analog for the robust RCPB lines of concern here.2 Per NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, thermal fatigue, stress corrosion cracking, and mechanical fatigue are the degradation mechanisms that most significantly contribute to LOCA.3 These mechanisms are precluded by the design of the RCPB components outside containment and they are detectable via the augmented inspection requirements should they arise. 1 NUREG/CR-5750, Appendix J: "The total world experience includes no reported large or medium [2 - 6 inch inner diameter] pipe break LOCAs in about 8,000 worldwide reactor calendar years of operation. No small [less than 2 inch inner diameter] pipe break LOCAs were reported in the total U.S. operating experience (about 2,100 reactor calendar years)." 2 NUREG-1829: "The major piping contributors for PWR [pressurized water reactor] Category 1 and 2 LOCAs are, as was the case for the BWR [boiling water reactor] plants, the instrument and drain lines." Also, "For PWRs, non-piping components are the dominant contributor to Category 1 and 2 LOCA frequencies due to SGTRs [steam generator tube ruptures] and CRDM [control rod drive mechanism] cracking." 3 NUREG/CR-5750, Appendix J: "Mechanisms responsible for degrading small diameter primary piping in PWRs include IGSCC [intergranular stress corrosion cracking] and other forms of stress corrosion cracking, thermal fatigue (PWR only), compression fitting failures in instrument lines, and vibration fatigue." And, "In small diameter piping, vibratory fatigue and compression fitting failures appear to occur most frequently." NuScale Nonproprietary

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 3 of 8 NuScale Nonproprietary Despite a very low frequency of occurrence, should an unisolable failure outside the CNV occur, the likelihood of successful mitigation is high. ((

}}2(a),(c)

Emergency Core Cooling System The ECCS valve flanges are analogous to other RPV appurtenances. Due to the robustness of the design code, quality assurance, and inspection requirements, ruptures of the RPV are historically excluded from design-basis events based on low likelihood of occurrence. Application of BTP 3-4, which requires augmented inspections and the use of more conservative stress limits for design, to the ECCS valve flange connections further reduces the likelihood of rupture in the US460 design. What are the Consequences? Chemical and Volume Control System ((

}}2(a),(c)

Emergency Core Cooling System The consequences of an ECCS valve rupture are ((

}}2(a),(c) 

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 4 of 8 NuScale Nonproprietary What could go right? Chemical and Volume Control System To complete the risk triplet discussion, an assessment of What could go right is provided. The objective of the assessment is to provide a look at the design from a holistic perspective of safety. The assessment considers: component design considerations inspection requirements stress analysis results insights from the US460 Probabilistic Risk Assessment controls on equipment available operator indications for an abnormality NuScale reduced the assembly between the CNV and CIVs to components and welds, with the CIVs located as close to the CNV as practical. The assembly measures approximately ((

}}2(a),(c), ECI from the base of the safe-end to the weld between the containment isolation test fixture (CITF) and the CIVs. Figure 1, Typical Assembly, depicts the assembly with its welds labeled. The purpose of this configuration is (( 
}}2(a),(c)

Figure 1: Typical Assembly

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 5 of 8 NuScale Nonproprietary In addition to the reduction to components and welds, NuScale has designed the welded connections to meet strict stress criteria for the purpose of further reducing the potential for a break in the region between the CNV and CIVs. Preliminary stress analysis for each of these locations show ((

}}2(a),(c) Tables 1, 2, and 3 provide the ASME code requirements applied to the welds, preliminary cumulative usage factor (CUF) results, and preliminary stress results, respectively.

Table 1: Application of Code at Connections Weld Location (Figure 1) ASME Component Type ASME Class (Subsection) Stress Analysis Cumulative Usage Factor Inservice Inspection CITF to CIV (B) Piping(1) Class 1 (NB) Section III, Subsection NB and Mandatory Appendix XIII, Paragraph XIII-3420 BTP 3-4 B.1.(ii)(b) ASME BPVC Section XI inspections augmented by BTP 3-4 B.1.(ii)(7) Safe-end to CITF (C) Piping(1) Class 1 (NB) Section III, Subsection NB and Mandatory Appendix XIII, Paragraph XIII-3420 BTP 3-4 B.1.(ii)(b) ASME BPVC Section XI inspections augmented by BTP 3-4 B.1.(ii)(7) Safe-end to Nozzle (D) Vessel(2) Class 1 (NB) Section III, Subsection NB and Mandatory Appendix XIII, Paragraph XIII-3450 BTP 3-4 B.1.(ii)(b) ASME BPVC Section XI inspections augmented by BTP 3-4 B.1.(ii)(7) Note 1: Locations are break exclusion zones and therefore have additional conservative requirements. Note 2: Location belongs to the CNV and meets ASME requirements for metal containment. The application of BTP 3-4 requirements highlights design requirements beyond traditional design requirements.

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 6 of 8 NuScale Nonproprietary Table 2: Preliminary Margin Results for Welds for Branch Technical Position 3-4 (( }}2(a),(c) Table 3: Preliminary Stress Margin Results for Welds (( }}2(a),(c) Welded connections located between the CNV and CIVs receive 100 percent volumetric inspections on a frequency of once per ISI interval (10 years). The volumetric inspection meets the inspection requirements of BTP 3-4. For the US460, this translates to 24 separate inspections (four locations and six NuScale Power Modules [NPMs]) over a ten-year period. In other words, there is an average of at least two representative inspections per year with six NPMs in operation. The number of NPMs (significant number of inspections) at a single plant ensures if a degradation at a location is detected at one NPM, it is expected and reasonable for the owner to take actions to remediate. The ASME Class 1 assembly configuration connecting the CNV and the CIVs includes a safe-end, CITF, and the CIVs. Prior to service, the CITF and CIVs receive a pressure test with a test pressure of 1.5 times cold working pressure. This test is required for all production valves prior to installation, and demonstrates the integrity of the valve body. Once installed, the RCPB is

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 7 of 8 NuScale Nonproprietary hydrostatically tested in accordance with ASME code. The combination of inspections and examinations for the assembly provides reasonable assurance of integrity. NuScales Probabilistic Risk Assessment considers pipe breaks outside of the CNV. While the model ((

}}2(a),(c)

For a CVCS injection line pipe break outside containment (including the CVCS injection line and pressurizer spray line), ((

}}2(a),(c) with either a safety-related passive success path or a nonsafety-related active success path. The safety-related success path requires a reactor trip, successful actuation of ECCS, and one train of DHRS. The nonsafety-related success path incudes reactor trip (safety-related), one reactor vent valve (safety-related), one reactor recirculation valve (safety-related), and makeup through the containment flooding and drain system (nonsafety-related). The success paths are the same for a CVCS discharge line break outside of containment.

Technical specifications (TS) and programmatic requirements (e.g., inservice testing, maintenance rule) ensure that safety-related components are reliable and capable of mitigating a break at the subject locations. Specifically, the module protection system is covered by TS 3.3.1, ECCS by TS 3.5.1, and DHRS by TS 3.5.2. In addition to the reduction of the assembly to components and welds, robustness of the components and welds, margin to stress limits, and the ability to mitigate the event, instrumentation and surveillances required by technical specifications provide early indication of a small leak. Technical Specification 3.4.5 requires performance of an RCS water inventory balance and primary to secondary leakage verification once per 72 hours (Surveillance Requirements 3.4.5.1 and 3.4.5.2). Under-the-bioshield temperature monitoring provides indication of a temperature change due to a leak. Abnormal frequency of makeup to the reactor from CVCS could also alert operators to the presence of a leak. Emergency Core Cooling System Overall, sensitivity studies show that ((

}}2(a),(c) 

NuScale Loss-of-Coolant Accident Safety Case December 18, 2024 Page 8 of 8 NuScale Nonproprietary ((

}}2(a),(c)

There are four bolted ECCS valve flange connections designed to ASME BPVC Section III, Subsection NB requirements. In addition, BTP 3-4 B.1.(ii) is applied for more conservative design stress and fatigue limits and additional ISI requirements. High strength bolting is required per ASME requirements with fatigue strength reduction factors not less than 4.0. Bolts require ultrasonic examination at service intervals. A small leak or crack at the locations (as opposed to full connection area opening) is also not a safety concern because: The limiting acceptance criterion (i.e., MCHFR) for these types of events based on sensitivity studies is less limiting for smaller break sizes. The design includes containment leakage detection. Leakage into containment is monitored by technical specifications. Containment pressure increase leads to automatic protective system actuations. Conclusion The NuScale US460 is designed to eliminate, to the extent practical, design-basis events, and to mitigate those that remain with simplified, passive safety systems. It is reasonable to conclude that the design is safe based on the exclusion of traditional reactor coolant pumps and large-bore RCS piping; the limited RCPB piping and welds; the application of robust design and inspection criteria; the use of simple, passive safety systems to mitigate accident sequences; the controls on safety-related equipment; the availability of operator indication; and the fact that ((

}}2(a),(c) for a postulated break at the ECCS valve flanges or between the CNV and CIVs for CVCS. 

LO-175581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Safety Case Updated Markup (Chapter 19), Nonproprietary

NuScale Final Safety Analysis Report Probabilistic Risk Assessment NuScale US460 SDAA 19.1-162 Draft Revision 2 SRP 19.0 AC II.8 SRP 19.0 AC II.14 SCP-2623, SCP-3075 Pending SCP2-4065 Table 19.1-28: Design-Specific Sources of Level 2 Model Uncertainty Uncertainty Source Description Level 2 Assumption Effect on Model Level 2 Analysis Large release definition Definition and modeling of a large release. The failure of containment isolation is assumed to result in a large release and there is no credit for mitigation (e.g., deposition). This is a bounding assumption; sequences with a failure of containment isolation are included in the frequency of a large release. Level 2 physical phenomena Susceptibility of the design to the typical severe accident phenomena that challenge containment, including hydrogen combustion, steam explosion, high pressure melt ejection, containment pressurization from a LOCA blowdown, overpressure. Based on design-specific analysis, the design is not susceptible to the typical severe accident phenomena. Severe accidents do not generate enough steam or hydrogen to pose a threat, the design pressure of the CNV is high, and immersion in the reactor pool is an effective heat removal mechanism. The containment event tree is limited to failures of containment isolation and induced SGTFs. This is judged not to be a significant source of model uncertainty because of CNV immersion in the reactor pool, which contains sufficient water inventory to cool modules for an extended period under adverse conditions. Unisolable failure of the reactor coolant pressure boundary outside of the CNV Potential failure of welds between the CNV and CIVs. The low likelihood of weld failures between the CNV and the CIVs for CVCS, combined with the capability to detect and requirements to respond to leaks, minimize the impacts on LRF. Because the assessment of this failure rate indicates a low likelihood and there are leak identification and response requirements (e.g., under-the-bioshield temperature indications, operator monitoring, Technical Specification 3.4.5), the impact on LRF is minimized and PRA insights are unchanged. Consequences of similar events (i.e., an unisolated CVCS break downstream of the CIVs) are already modeled in the PRA.

LO-175581 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-175582

AF-175582 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 report reveals distinguishing position of NuScale LOCA Break Safety Case. NuScale has performed significant research and evaluation to develop a basis for this process and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled, NuScale LOCA Break Safety Case. 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-175582 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 NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on January 31, 2025. Mark W. Shaver}}