ML24146A001
| ML24146A001 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 05/25/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24146A001 (6) | |
Text
From:
Getachew Tesfaye Sent:
Saturday, May 25, 2024 9:25 AM To:
Request for Additional Information Cc:
Prosanta Chowdhury; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Fairbanks, Elisa; NuScale-SDA-720RAIsPEm Resource
Subject:
NuScale SDAA Section 17.4 - Request for Additional Information No. 026 (RAI-10199-R1)
Attachments:
SECTION 17.4 - RAI-10199-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:
35 Mail Envelope Properties (BY5PR09MB5682AA0E2F16893541A449488CF62)
Subject:
NuScale SDAA Section 17.4 - Request for Additional Information No. 026 (RAI-10199-R1)
Sent Date:
5/25/2024 9:25:26 AM Received Date:
5/25/2024 9:25:30 AM 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@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 584 5/25/2024 9:25:30 AM SECTION 17.4 - RAI-10199-R1-FINAL.pdf 179198 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 026 (RAI - 10199-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 17, QUALITY ASSURANCE AND RELIABILITY ASSURANCE SECTION 17.4, RELIABILITY ASSURANCE PROGRAM ISSUE DATE: 05/25/2024 Question 17.4-11
Background
By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR), Chapter 17, Quality Assurance and Reliability Assurance, Revision 1 (Agencywide Documents Access and Management System Accession No. ML23304A371) of the NuScale Standard Design Approval Application (SDAA) for its US460 standard plant design. The applicant submitted the US460 plant SDAA in accordance with the requirements of Title 10 of the 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 provided in FSAR Chapter 17 of the SDAA and has determined that additional information is required to complete its review.
Question 17.4-11 Regulatory Basis 10 CFR 52.137(a)(9) requires that an application must contain a final safety analysis report which must include, in part, the following: for applications for light-watercooled nuclear power plants, an evaluation of the standard plant design against the Standard Review Plan (SRP) revision in effect 6 months before the docket date of the application. The evaluation required by this section shall include an identification and description of all differences in design features, analytical techniques, and procedural measures proposed for the design and those corresponding features, techniques, and measures given in the SRP acceptance criteria.
The purpose, scope, and criteria of the Reliability Assurance Program (RAP), as discussed in section 17.4 of NUREG-0800, are established in:
SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Nonsafety Systems in Passive Plant Designs, dated March 28, 1994 (ML003708068), and associated Staff Requirements Memorandum (SRM), dated June 30, 1994 (ML003708098), and SECY-95-132, Policy and Technical Issues Associated with the Regulatory Treatment of Nonsafety Systems (RTNSS) in Passive Plant Designs
2 (SECY94084), dated May 22, 1995 (ML003708005), and associated SRM, dated June 28, 1995 (ML003708019).
Issue The NuScale US460 (SDA) design is a passive advanced light-water reactor (ALWR), which is covered by the regulations and Commission policies identified above. The SRM to SECY 084 states that the purposes of the RAP program are to provide reasonable assurance that (1) an ALWR is designed, constructed, and operated in a manner that is consistent with the assumptions and risk insights for these risk-significant structures, systems, and components (SSCs), (2) the risk-significant SSCs do not degrade to an unacceptable level during plant operations, (3) the frequency of transients that challenge ALWR SSCs are minimized, and (4) these SSCs function reliably when challenged.
In SDAA FSAR Section 17.4.3.1, Structures, Systems, and Components Classification and Categorization Process, the applicant described the overall SSC classification process. The applicant stated, System functions and the SSC that perform those functions are evaluated for risk-significance based on a consideration of probabilistic, deterministic, and other methods of analysis, including industry operating experience, expert panel reviews, and severe accident evaluations.
In SDAA FSAR Section 17.4.3.2, Identification of Design Reliability Assurance Program Structures, Systems, and Components, the applicant stated, Concurrence by the expert panel constitutes the final classification of the SSC. The applicant also stated, The risk-significance classification for safety-related equipment is the default classification unless the PRA determined that the SSC functionalities are not risk-significant.
In SDAA FSAR Figure 17.4-1, Structures, Systems, and Components within the Scope of the Reliability Assurance Program, the applicant illustrated its process for determining the risk significance of SSCs. This figure explicitly identified operating experience, PRA and severe accident insights and assumptions, defense in depth, and systems interactions as additional considerations when the expert panel considers the safety and risk categorizations.
In SDAA FSAR Section 17.4.3.1, the applicant also described that it uses the approach approved in Licensing Topical Report TR-0515-13952-NP-A, Revision 0, Risk Significance Determination (ML16284A016). In the Final Safety Evaluation Report for Licensing Topical Report TR-0515-13952-NP-A, dated July 13, 2016 (ML16181A218), the NRC staff concluded that the methods described in TR-0515-13952-NP-A are acceptable for identifying SSCs as candidates for risk significance in a NuScale design PRA, subject to the conditions and limitations provided. Condition and limitation 2 states, in part:
In keeping with NRC policy on risk-informed regulation, the ultimate determination of risk significance shall be based on the specific application, with appropriate consideration of uncertainties, sensitivities, traditional engineering evaluations and regulations, and maintaining sufficient defense-in-depth and safety margin. As such, PRA risk insights shall be considered along with deterministic approaches and defense-in-depth concepts such that the user is implementing a risk-informed rather than a solely risk-based approach.
3 The NRC staff used the guidance in SRP Section 17.4, Revision 1, to conduct its review of the applicants design reliability assurance (D-RAP) program. The NRC staff focused on SSCs with design changes compared to the certified US600 design to provide a more effective and efficient review and audited the system function reports for the steam generator system, control rod drive system, decay heat removal system, boron addition system, emergency core cooling system, containment system, and reactor coolant system. During the audit, the NRC staff was unable to verify through the review of records an adequate implementation of the D-RAP process; specifically, NRC staff was unable to verify demonstrations, with supporting documentation, of what specific deterministic and defense-in-depth considerations were considered by the D-RAP expert panel and how these inputs were dispositioned by the D-RAP expert panel in its risk significance decisions for the example SSCs and functions for the SDA design. Based on the NRC staffs review of the SDAA FSAR, the system function reports, and documentation in the regulatory audit, the NRC staff is unable to find evidence to conclude that the applicant implemented a risk-informed process for SSC classification. Specifically, the NRC staff is unable to conclude that deterministic considerations and defense-in-depth were appropriately considered in the risk significance determination of SSCs, especially in cases where there are no apparent design differences for the SSCs between the US460 (for which the SSCs were deemed not risk significant) and the certified US600 designs (for which the SSCs were deemed risk significant).
Additionally, in SDAA FSAR Table 19.155, Shared System Hazard Analysis, the applicant stated, [t]he loss of the backup power supply system (BPSS) would reduce defense-in-depth of the station in response to a loss of offsite power event. More than 25 percent of the internal events CDF caused by losses of offsite power is mitigated by the two backup diesel generators (BDGs) without the need to initiate ECCS. The staff also notes that the two BDGs support all six NPMs in the US460 design, compounding the impact of the reliability of the BDGs. Yet, the BDGs are not scoped into D-RAP.
The NRC staff notes that the thresholds for candidate risk significance from the PRA are different between the SDA and certified US600 designs and that these thresholds in the SDAA already account for the low absolute risk of the SDA design compared to legacy plants.
During the audit, the NRC staff requested the primary system or plant design changes in the SDA design that led to a change in categorization of the control rod drive, containment, and steam generator systems. In its response, the applicant described that the differences in risk significance classifications from the certified US600 design to the SDA design were not necessarily the result of a design change to the SSC, but instead reflective of performing the evaluations with updated information and no longer assuming a default classification of risk significant for safety-related equipment. The NRC staff determined that this response contradicts information provided in SDAA FSAR Section 17.4.3.2.
Information Requested To support the NRC staffs finding against 10 CFR 52.137(a)(9) on the SDAAs conformance with the Commissions direction on D-RAP, NuScale is requested to:
4
- 1. Confirm that the SSC classification process was performed in accordance with SDAA FSAR Section 17.4.3.2, which states, The risk-significance classification for safety-related equipment is the default classification unless the PRA determined that the SSC functionalities are not risk-significant. If the default classification for safety-related equipment is not risk significant, clarify the SSC classification process, with justification, and provide an FSAR markup to reflect that process.
- 2. For the control rod drive and steam generator systems (i.e., the systems whose functions were categorized as risk significant in the certified US600 design and not risk significant in the SDA design), provide an FSAR markup of Table 17.4-1 that classifies the functions and required SSCs as risk significant. Alternately, justify that the system function categorization and subsequent SSC classification is risk informed. As part of the justification, for each of the above-mentioned systems, (i) describe the specific deterministic and defense-in-depth considerations that were evaluated by the D-RAP expert panel and how these considerations were dispositioned and (ii) discuss the leading system or plant design changes that drove the determination by PRA that the system functions were not risk significant.
- 3. The reactor coolant pressure boundary (RCPB) is essential to defense in depth in the SDA design, especially during normal operation when the emergency core cooling system (ECCS) is not activated, which is the dominant operational configuration. Despite its importance, maintaining RCPB is not identified as risk significant for any system in SDAA FSAR Table 17.4-1. Given the importance of maintaining the RCPB to defense in depth, especially during normal operation when ECCS is not activated, provide an FSAR markup of Table 17.4-1 that classifies this function and the required SSCs as risk significant.
- 4. For the BPSS, whose loss, according to the FSAR, would reduce defense in depth, justify that the system function categorization and subsequent SSC classification is risk-informed, and not solely risk-based, given the successful operation of the BDGs is necessary following an extended loss of AC power to prevent unnecessary ECCS actuation in the SDA design and mitigate more than 25 percent of the CDF. As part of the justification, describe the specific deterministic and defense-in-depth considerations that were evaluated by the D-RAP expert panel for the BPSS and how these considerations were dispositioned.