ML24290A192
| ML24290A192 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 10/16/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24290A192 (5) | |
Text
From:
Getachew Tesfaye Sent:
Wednesday, October 16, 2024 7:36 PM To:
Request for Additional Information Cc:
Mahmoud -MJ-Jardaneh; Griffith, Thomas; Osborn, Jim; NuScale-SDA-720RAIsPEm Resource
Subject:
Nonproprietary - NuScale SDAA Section 6.2 - Request for Additional Information No. 039 (RAI-10359-R1)
Attachments:
SECTION 6.2 - RAI-10359-R1-FINAL NON-PROPRIETARY.pdf Attached please find NRC staffs nonproprietary 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). The encrypted proprietary version will be submitted in a separate email.
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:
39 Mail Envelope Properties (BY5PR09MB56829BADF7DCF84E8C27FF678C462)
Subject:
Nonproprietary - NuScale SDAA Section 6.2 - Request for Additional Information No. 039 (RAI-10359-R1)
Sent Date:
10/16/2024 7:36:12 PM Received Date:
10/16/2024 7:36:17 PM From:
Getachew Tesfaye Created By:
Getachew.Tesfaye@nrc.gov Recipients:
"Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>
Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>
Tracking Status: None "Osborn, Jim" <josborn@nuscalepower.com>
Tracking Status: None "NuScale-SDA-720RAIsPEm Resource" <NuScale-SDA-720RAIsPEm.Resource@nrc.gov>
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 673 10/16/2024 7:36:17 PM SECTION 6.2 - RAI-10359-R1-FINAL NON-PROPRIETARY.pdf 131874 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 039 (RAI-10359-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 6, ENGINEERED SAFETY FEATURES SECTION 6.2.1, CONTAINMENT FUNCTIONAL DESIGN &
SECTION 6.2.2, CONTAINMENT HEAT REMOVAL ISSUE DATE: 10/16/2024
=
Background===
By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR), Chapter 6, Engineered Safety Features, Revision 1 (Agencywide Documents Access and Management System Accession No. ML23304A345), 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 FSAR Chapter 6 of the SDAA, and LOCA EM LTR, Revision 3 for the containment response analysis methodology (CRAM) and determined that additional information is required to complete its review.
Question 6.2.1-1 Regulatory Basis GDC 16, Containment Design, as it relates to the reactor containment and associated systems being designed to assure that containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
GDC 38, Containment Heat Removal, as it relates to the containment heat removal system(s) (CHRS) function to rapidly reduce the containment pressure and temperature following any LOCA and maintain them at acceptably low levels.
GDC 50, Containment Design Basis, as it relates to the reactor containment structure and associated heat removal system(s) being designed so that the containment structure and its internal compartments can accommodate the calculated pressure and temperature conditions resulting from any LOCA without exceeding the design leakage rate and with sufficient margin.
Issues (a) The LOCA EM LTR, Revision 3, Section 9.6.1, presents the impact of three containment vessel (CNV)-reactor pressure vessel (RPV) nodalization schemes, summarized in Table 9-7, on the NPM-20 LOCA figures of merit (FOMs) for the reactor coolant system (RCS) injection line and high point vent line breaks, with no DHRS operation credited. No nodalization sensitivity results are presented for the limiting containment design basis accident (DBA), i.e.,
RCS discharge line (DL) break. ((
2. As opposed to US600, the US460 design relies on DHRS operation for the containment DBA mitigation, which would lead to an increasing level of CNV pressurization sensitivity to DHRS performance as the break size becomes smaller and the ECCS actuation is delayed. Therefore, during the audit, the staff requested information about the containment vessel and reactor pool nodalization sensitivity and natural convection modeling assumptions for the US460 design. The staff had requested the information as the reactor pool heat up and thermal stratification could degrade the DHRS capacity as the DBA progresses, which would potentially lead to a higher CNV pressurization under delayed ECCS actuation, especially for small break LOCA. (( }}. NuScale provided no information for the US460 containment and pool nodalization sensitivity to address the staff concerns about the DHRS performance degradation due to pool heat up and thermal stratification and its impact on containment pressurization accompanied by delayed ECCS actuation, especially toward the smaller break end of the spectrum. None of the information provided in the audit or available in the SDAA supports a bounding minimum DHRS capacity for the limiting containment analyses. The staff needs additional information to ensure that there is no CNV response that could potentially challenge the limiting CNV design basis event for the NPM-20 containment design. (b) During the audit, the staff also requested NuScale to provide the description and assumptions made for the reactor pool bay corresponding to an individual NPM, and the nodal implementation of the natural convection correlation on the CNV external surface. (( }}. (c) The staff compared the wall condensation and interphase heat and mass transfer correlations between the LOCA EM TR, Revision 2 and Revision 3. (( }}.
3 Information Requested (a) NuScale is requested to provide the LOCA FOM results for both the CNV and reactor pool nodalization sensitivities, or demonstrate that the containment thermal-hydraulic response is not sensitive to the CNV and reactor pool nodalizations, (( }}. The staff requests the analyzed containment transient results for 72 hours to ensure that the CNV pressure that is reduced to half within 24 hours would be maintained at acceptably low level (as required by GDC 38) despite the reactor pool heat up and DHRS performance degradation. (b) The staff requests a justification for the natural convection modeling of the CNV and DHRS by using an overall average heat transfer coefficient based on the surface height, for pool nodalization to be conservative with respect to the containment pressurization and DHRS capacity. Please justify this approach compared to accounting for natural convection from the CNV and DHRS to the pool driven by the local natural heat transfer coefficient for each node based on the local wall and local bulk pool temperatures. (c) (( }}.}}