ML25091A094
| ML25091A094 | |
| Person / Time | |
|---|---|
| Issue date: | 04/02/2025 |
| From: | Halnon G Advisory Committee on Reactor Safeguards |
| To: | Walter Kirchner Advisory Committee on Reactor Safeguards |
| References | |
| Download: ML25091A094 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, DC 20555 - 0001 April 2, 2025 MEMORANDUM TO:
Walter L. Kirchner, Chairman NuScale Subcommittee Advisory Committee on Reactor Safeguards FROM:
Gregory Halnon, Member NuScale Subcommittee Advisory Committee on Reactor Safeguards
SUBJECT:
INPUT FOR ACRS REVIEW OF NUSCALE POWER, LLC, STANDARD DESIGN APPROVAL APPLICATION - SAFETY EVALUATION REPORT FOR CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS, COMPONENTS AND EQUIPMENT In response to the Subcommittees request, I have reviewed the NRC staffs advance safety evaluation report (SER) provided to support ACRS review of the NuScale standard design approval application (SDAA), and the associated section of the applicants submittal for Chapter 3, Design of Structures, Systems, Components and Equipment, Revision 1. I also received input from consultants Dennis Bley and Stephen Schultz. The following is my recommended course of action concerning further review of this chapter and the staffs associated safety evaluation.
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Background===
Chapter 3 of the NuScale Final Safety Analysis Report (FSAR) documents the analytical methods, testing procedures, tests and analyses that the applicant used to ensure the structural and functional integrity of the piping systems, mechanical equipment, reactor vessel, reactor internals, and their supports under static and dynamic loadings, including those caused by normal operation and postulated events. The SER documents the staffs review of this FSAR Chapter. The FSAR Chapter 3 incorporates, by reference, Licensing Technical Report, TR-121517-P, NuScale Power Module Short-Term Transient Analysis, Revision 0. This report supplements descriptions and results from the dynamic analyses performed to evaluate structural response of the NuScale Power Module (NPM). Short-term transients are events caused by the failure or actuation of piping and valves and include high-energy line breaks. These events result in system internal pressure waves or asymmetric cavity pressurization waves exterior to the pipe break or valve outlet. This chapter also provides information on the seismic design, including details on the seismic analyses of the two Seismic Category I structural portions of the Control Room Building and Reactor Building (RXB).
SER Summary The NRC staffs SER for this chapter provides a concise narrative of the differences between the previously approved certified design for the US600 and the submitted US460 SDAA. In the
sections reviewed, there were only a few significant differences, and all were found to be acceptable and had, if applicable, relevant requirements to be verified or completed during a combined license application (termed COL items). Accordingly, the staff confirmed the COL items were adequate to ensure full compliance with guidance and regulations.
The seismic design of the NPM includes the necessary parameters for a complete analysis, primarily derived from guidance provided in the suite of seismic-related regulatory guides and the Design Specific Review Standard. Additionally, where appropriate, a variety of industry standards and related regulatory information are applied. Seismic information presented was found to be acceptable for this stage of the design. Where site specific information would need to be evaluated, the staff confirmed the COL items were adequate to ensure full compliance with guidance and regulations. Analytical methods and modeling of the effects on structures, systems, and components were all found to be reasonable and acceptable according to established guidance and standards. Seismic instrumentation was adequately described, and an associated COL item will ensure the maintenance and post-earthquake evaluations are appropriate.
The staff confirmed engineering documents were acceptable during an audit of the design; however, some analytical calculations were not final. The staff verified that there are approved QA-pedigreed processes to track their completion (corrective action program or verified during Inspections, Tests, Analyses, and Acceptance Criteria performance).
Vibration related analyses and test methods are covered in the Comprehensive Vibration Assessment Program (CVAP). The staff found the approach complete and adequate to ensure the safety case is acceptable.
No open items were identified.
Discussion The information presented is consistent with regulatory guidance and requirements. No deficiencies were identified where additional information was needed to confirm adequacy or augment the descriptions, analyses, or conclusions in the SDA chapter. Where assumptions were made, they were conservative and met or exceeded guidance expectations.
The Certified Seismic Design Response Spectra (CSDRS) and CSDRS-HF (CSDRS High Frequency) was compared against the preliminary data presented in the NRC Memorandum "Support Document for Screening and Prioritization Results Regarding Seismic Hazard Re-Evaluations for Operating Reactors in the Central and Eastern United States." The CSDRS and CSDRS-HF provide a reasonable envelope for most site soil conditions. There was one case where the staff did not document the acceptability of strong-motion time history being less than six seconds. The FSAR narrative justifies this as acceptable due to a longer period of strong motion; however, the staff SER did not directly address this anomaly. When this feedback was provided to the staff, they revised the SER to include the necessary review information and justification of acceptability.
The overall approach for the Density Wave Oscillation safety case is anchored on three pillars:
real-time monitoring, analyses supported by experimental testing programs, and physical inspections. The CVAP, ASME Section XI ISI, and Steam Generator (SG) Program are important programs underlying the US460 NPM SG design and operation. The CVAP provides for extensive instrumentation upon initial operations. It will use classic system parameter measuring
instrumentation (flow, pressure, temperature) and pressure-wave transducers to detect onsets of fluid elastic instability, flutter/gallop, leakage flow, turbulent buffering, and vortex shedding phenomena. Using test facility data, there is high confidence in the ability to show vibration is within expected and acceptable values. Additionally, the established DWO operating curve in conjunction with SG tube inlet flow restrictors will provide a practical operating envelope and appropriately track potential DWO hours ensuring this phenomenon does not cause SG tube failure.
In Regulatory Guide 1.20, Section C.1.1, it defines a prototype as a configuration of reactor internals, or a single component that, because of the arrangement, design, size, or operating conditions, represents a first-of-a-kind or unique design for which no previous valid prototype can be referenced. Since this SDA is applied to a six-pack (six NPMs), NuScale provided the clarification that, until the first modules vibration data is analyzed as acceptable, all modules in operation will be treated as prototypes per the regulatory guide and instrumented accordingly. After the NRC staff accepts the prototype design and the CVAP is completed with no adverse vibration or excessive loading, the initial NPM may be classified as a valid prototype. As a valid prototype, it may be referenced in subsequent applications, as appropriate, subject to its restrictions and provisions. This is an important aspect of the CVAP because non-prototype modules will credit the acceptability of the design, related to vibration, based on the CVAP prototype.
The following narrative is not specific to any deficiency or to Chapter 3 in the SDAA. This issue is for awareness during design reviews and generally transcends the SDA process. An important aspect of an SDA is the incremental nature of the process to achieve a first-of-a-kind (FOAK) reference plant. The Advance Act of 2024, generally, has an expectation that licensing nth-of-a-kind (NOAK) reactor plants will involve an expedited review and approval process due to the standardization of key design features and an integrated plant design. The license application for an FOAK plant, using either a Title 10 of the Code of Federal Regulations Part 50 construction or operating license, or a Part 52 combined license application (COLA), will provide a complete detailed design that enables NOAK-type reviews and approvals during a subsequent site-specific combined license application (SCOLA). Design detail attributes that need to be carried from an essentially complete design presented in an SDAA to an COLA is important to ensure confidence of adequate safety in an integrated fashion, not just on a system-by-system basis. As mentioned, the application for the FOAK is truly the reference plant, so SDA items to be verified and completed during the COLA stage should be clear and explicit. As examples within SDAA Chapter 3, some important design features to be handled by COL items include:
- 1. In the case of high-energy and moderate-energy piping systems not associated with the containment vessel, the pipe rupture hazards analysis is preliminary and the evaluations are assumed to bound final configurations. Confirmation of acceptable final configurations and hazards analysis are left to the Inspections, Tests, Analyses, and Acceptance Criteria and COL Item 3.6-1.
- 2. The standard design leaves final design details of equipment subject to flood protection to the COLA. These details include post-accident flood protection features such as location, flood levels, and flood protection procedures. The COLA applicant must confirm that potential site-specific flooding sources do not negate these protection features. For example, the safety-related Augmented DC Power System (battery rooms) and their post-accident internal flood protection must be finalized and confirmed (COL Item 3.4-1 and 3.4-2).
- 3. Turbine missiles are shown to not penetrate the critical walls of the RXB in most cases.
The one case (region of the shortest load path) is due to an unfavorable orientation of the turbine to the RXB. This safety feature is dependent on the location and orientation of the turbine to the RXB walls. Given some site-specific turbine facilities may be repurposed or there may be limited real estate availability, turbine orientation could be challenging and its safety impact will need to be evaluated during the COLA stage (COL Item 3.5-1).
There are also some assumptions made for items of lesser safety significance that may need to be addressed. For example, the determination of the testing frequency for safety valves for the SDA is based on presumed data from operation of six modules. If a future owner decides fewer modules are to be constructed and operated on a site, the testing frequency assumptions must be revisited to ensure they remain conservative and valid. Similar considerations may be required for this configuration to be reviewed as appropriate during the COLA stage.
For NuScale, the benefits of the commonality of the US460 SMR commercial fleet should be established and carried through to a quality first reference COLA (RCOLA), with successful licensing review interaction to gain timely approval of the operating license. Once this is mastered for the FOAK, the choice of leveraging this standardization during any NOAK licensing process should be apparent and will be important to achieving the licensing efficiency and effectiveness goals of the ADVANCE Act of 2024. Conversely, if subsequent owners do not desire or cannot follow the RCOLA with an almost identical specific or SCOLA, the difficulties experienced with the divergence of standardization in the current U.S. light-water reactor fleet may be repeated.
Recommendation No further review of Chapter 3 is warranted. The more generic discussion regarding efficient FOAK and NOAK approvals in consideration of the Advance Act of 2024 should be included in our next lessons learned report on recent advanced design reviews.
References
- 1. U. S. Nuclear Regulatory Commission, Safety Evaluation of NuScale SDAA Chapter 3, Design of Structures, Systems, Components and Equipment, October 7, 2024 (ADAMS Accession No. ML24289A089).
- 2. NuScale Power, LLC, Standard Design Approval Application, Part 2, Chapter 3, Design of Structures, Systems, Components and Equipment, Revision 1, October 31, 2023 (ADAMS Accession Nos. ML23304A321 (Public) ML23304A322 (Non-Public)).
- 3. 118th Congress of the United States, An Act, Division B - Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, January 3, 2024.
- 4. NuScale Power, LLC, Licensing Technical Report, TR-121517, NuScale Power Module Short-Term Transient Analysis, Revision 0, October 31, 2023 (ADAMS Accession Nos.
ML23304A333 (Public) ML23304A334 (Non-Public)).
April 2, 2025
SUBJECT:
INPUT FOR ACRS REVIEW OF NUSCALE POWER, LLC, STANDARD DESIGN APPROVAL APPLICATION -SAFETY EVALUATION REPORT FOR CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS, COMPONENTS AND EQUIPMENT Package Accession No: ML25091A091 Accession No: ML25091A094 Publicly Available (Y/N): Y Sensitive (Y/N): N If Sensitive, which category?
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NRC Users or ACRS only or See restricted distribution OFFICE ACRS SUNSI Review ACRS ACRS NAME MSnodderly MSnodderly LBurkhart GHalnon DATE 3/31/25 3/31/25 3/31/25 4/02/25 OFFICIAL RECORD COPY