ML24103A247
| ML24103A247 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 04/12/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24103A247 (5) | |
Text
From:
Getachew Tesfaye Sent:
Friday, April 12, 2024 5:41 PM To:
Request for Additional Information Cc:
Prosanta Chowdhury; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Fairbanks, Elisa; NuScale-SDA-720RAIsPEm Resource
Subject:
NuScale SDAA Section 3.6.2 - Request for Additional Information No. 023 (RAI-10177-R1)
Attachments:
SECTION 3.6.2 - RAI-10177-R1-FINAL.pdf Attached please find NRC staffs request for additional information (RAI) concerning the review of NuScale Standard Design Approval Application for its US460 standard plant design (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23306A033).
Please submit your technically correct and complete response by the agreed upon date to the NRC Document Control Desk.
If you have any questions, please do not hesitate to contact me.
Thank you, Getachew Tesfaye (He/Him)
Senior Project Manager NRC/NRR/DNRL/NRLB 301-415-8013
Hearing Identifier:
NuScale_SDA720_RAI_Public Email Number:
32 Mail Envelope Properties (BY5PR09MB56821F69483A8FFD2B9F7A5E8C042)
Subject:
NuScale SDAA Section 3.6.2 - Request for Additional Information No. 023 (RAI-10177-R1)
Sent Date:
4/12/2024 5:40:47 PM Received Date:
4/12/2024 5:40:52 PM From:
Getachew Tesfaye Created By:
Getachew.Tesfaye@nrc.gov Recipients:
"Prosanta Chowdhury" <Prosanta.Chowdhury@nrc.gov>
Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>
Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>
Tracking Status: None "Fairbanks, Elisa" <EFairbanks@nuscalepower.com>
Tracking Status: None "NuScale-SDA-720RAIsPEm Resource"
<NuScale-SDA-720RAIsPEm.Resource@usnrc.onmicrosoft.com>
Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>
Tracking Status: None Post Office:
BY5PR09MB5682.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 584 4/12/2024 5:40:52 PM SECTION 3.6.2 - RAI-10177-R1-FINAL.pdf 146928 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 023 (RAI-10177-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS, COMPONENTS AND EQUIPMENT SECTION 3.6.2, DETERMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING ISSUE DATE: 04/12/2024
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Background===
By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR), Chapter 3, Design of Structures, Systems, Components and Equipment, Revision 1 (Agencywide Documents Access and Management System Accession No. ML23304A321), of the NuScale Standard Design Approval Application (SDAA) for its US460 standard plant design. The applicant submitted the US460 standard plant SDAA in accordance with the requirements of Title 10 Code of Federal Regulations (10 CFR)
Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, Subpart E, Standard Design Approvals. The NRC staff has reviewed the information in Chapter 3 of the SDAA and determined that additional information is required to complete its review.
Question 3.6.2-2 Regulatory Basis 10 CFR 50.55a Codes and standards requires that Class 1, 2, and 3 piping systems be designed to ASME Section III criteria.
10 CFR Part 50 Appendix-A, GDC 4, Environmental and dynamic effects design bases, states that structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
10 CFR 50.46, Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors, describes, in part, that emergency core cooling systems cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
2 Issue SRP Branch Technical Position (BTP) 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, provides guidance on postulated break locations and piping system design considerations.
In the Standard Design Approval (SDA) Pipe Rupture Hazards Analysis Technical Report (TR-121507-P), it is noted that containment system piping, which includes injection, discharge, Pressurizer (PZR) spray piping, and degasification piping, is classified as break exclusion and is required to meet applicable stress criteria of BTP 3-4 B.1.(ii). The containment system piping includes safety related ASME BPV Code Section III class 1 welded connections that constitute the reactor coolant pressure boundary. These welded containment penetration locations are important not only from 10 CFR part 50 Appendix A GDC 4, but are also extremely critical from 10 CFR Part 50.46 considerations.
Loss of fluid due to breaks in the welds at Containment Vessel (CNV) to CNV nozzle safe end, CNV nozzle safe end to Containment Isolation Valve (CIV) weld, or CIV outboard weld in the containment penetration area of Chemical and Volume Control System (CVCS) injection and discharge piping welded connections, Pressurizer (PZR) spray piping, and degasification line welded connections can result in significant consequences. Gross failure of welds at these locations would result in a LOCA that would challenge fuel integrity with concurrent containment bypass.
Breaks in these locations would also result in dynamic effects. BTP3-4 points out that subject to certain limitations, GDC 4 allows that dynamic effects associated with postulated pipe ruptures be excluded from the design basis when analyses reviewed and approved by the NRC demonstrate that the probability of fluid system piping rupture is extremely low under design-basis conditions. This can be satisfied, in part, by confirming that stress and cumulative usage factors (CUF) margins meet at least the BTP3-4 criteria using acceptable analysis methods, having appropriate augmented measures in materials, fabrication, and preservice and inservice inspections (PSI and ISI). The information provided in the application does not provide the staff sufficient information to make a regulatory finding. Specifically, a summary of the analyses and additional measures described above have not been provided as part of the application.
If breaks at these connections are not assumed to satisfy 10 CFR 50.46, then appropriate justification must be provided. The decrease in defense-in-depth due to not postulating losses of coolant accidents at these locations can be compensated, in part, by a high confidence in the structural integrity of the welded connections such that the probability of fluid system piping rupture is extremely low under design basis conditions are reviewed and approved by the NRC.
This can be satisfied, in part, by confirming that stress and cumulative usage factors (CUF) margins meet at least the BTP3-4 criteria using acceptable analysis methods, having appropriate augmented measures in materials, fabrication, and preservice and inservice inspections assuring the robustness of design.
Postponing completion of the preliminary analyses based on the current design until the ITAAC phase does not provide adequate assurance demonstrating the robustness of the design and margins for these welded connections. The staff does not consider regulatory commitments to complete the design evaluation and meet the BTP3-4 Break Exclusion area requirements and ASME BPV Code acceptance criteria through an ITAAC as an appropriate resolution for these
3 highly safety significant weld connections due to the far-reaching implications from 10 CFR Appendix A, GDC 4, and 10 CFR 50.46 perspectives.
Information Requested Provide a summary of preliminary stress analysis results (stresses and CUF values as applicable) for welded joints, including all welds between the CNV vessel nozzle to CIV outboard weld in the containment penetration area for the CVCS injection and discharge lines, PZR spray piping, and degasification lines, to demonstrate the robustness of the design and margins in the results. Describe additional measures that will be implemented for PSI and ISI beyond the ASME Section III and Section XI requirements for these locations. Include a markup of the FSAR, as appropriate, with the summary of significant assumptions, analysis results, and any measures taken to ensure the probability of fluid system piping rupture is extremely low under design-basis conditions.