ML24096B785
| ML24096B785 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 04/05/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24096B785 (5) | |
Text
From:
Getachew Tesfaye Sent:
Friday, April 5, 2024 8:57 PM To:
Request for Additional Information Cc:
Alina Schiller; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Sfowler@nuscalepower.com; NuScale-SDA-720RAIsPEm Resource
Subject:
NuScale SDAA Section 19.3 - Request for Additional Information No. 019 (RAI-10146-R1)
Attachments:
SECTION 19.3 - RAI-10146-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:
31 Mail Envelope Properties (BY5PR09MB5682219CA4037159D6BDA2558C022)
Subject:
NuScale SDAA Section 19.3 - Request for Additional Information No. 019 (RAI-10146-R1)
Sent Date:
4/5/2024 8:57:09 PM Received Date:
4/5/2024 8:57:13 PM From:
Getachew Tesfaye Created By:
Getachew.Tesfaye@nrc.gov Recipients:
"Alina Schiller" <Alina.Schiller@nrc.gov>
Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>
Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>
Tracking Status: None "Sfowler@nuscalepower.com" <sfowler@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 4/5/2024 8:57:13 PM SECTION 19.3 - RAI-10146-R1-FINAL.pdf 188429 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 019 (RAI-10146-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 19, PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT EVALUATION SECTION 19.3, REGULATORY TREATMENT OF NONSAFETY SYSTEMS ISSUE DATE: 04/05/2024
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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 19, "Probabilistic Risk Assessment and Severe Accident Evaluation," Revision 1 (Agencywide Documents Access and Management System Accession No. ML23304A385) 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 Chapter 19 of the SDAA and determined that additional information is required to complete its review.
Question 19.3-4 Regulatory Basis 10 CFR 52.137(a)(25) The application must contain a final safety analysis report that describes the design-specific probabilistic risk assessment and its results.
The scope, criteria, and process used to determine Regulatory Treatment of Nonsafety Systems (RTNSS) for the advanced passive plant designs are established in:
- 1. SECY-94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Nonsafety Systems in Passive Plant Designs, dated March 28, 1994 (Agencywide Documents Access and Management System Accession No. ML003708068) and associated Staff Requirements Memorandum (SRM), June 30, 1994 (ML003708098);
- 2. SECY-95-132, Policy and Technical Issues Associated with the Regulatory Treatment of Nonsafety Systems (RTNSS) in Passive Plant Designs, dated May 22, 1995 (ML003708005), and associated SRM, June 28, 1995 (ML003708019); and
- 3. SECY-96-128, Policy and Key Technical Issues Pertaining to the Westinghouse AP600 Standardized Passive Reactor Design, June 12, 1996 (ML003708224), and associated SRM, January 15, 1997 (ML003755486).
Issue The NuScale SDAA design is an advanced passive light water reactor, which is covered by the Commission policy identified above. The NuScale SDAA internal events full power core damage frequency (CDF) is 6x10-9/yr. For comparison, the GEH Economic Simplified Boiling Water Reactor (ESBWR) certified design is also an advanced passive design covered by the
2 Commission direction above. NUREG-1966, Final Safety Evaluation Report Related to the Certification of the Economic Simplified Boiling-Water Reactor Standard Design, Chapter 19, page 19-18 (ML14100A187), provides the ESBWR internal event full power CDF as 1.65x108/yr. The staff considers this risk result to be comparable to the NuScale SDAA CDF given that these are design probabilistic risk assessments (PRAs) with associated assumptions and uncertainties. The staff notes that the RTNSS initiating event guidance, as described in Standard Review Plan (SRP) 19.3 does not have a minimum baseline threshold on CDF. The staff has concluded that the Commission direction on RTNSS applies in its entirety to the NuScale SDAA design as it did to the design certification application (DCA) design. SDAA FSAR Section 19.3, where NuScale discusses how it meets the Commission policy on RTNSS without any exceptions, confirms the staffs conclusion.
The staff notes that the core damage risk profile for the US460 (SDA) design is significantly different from the US600 (DCA) design. In the US600 design, drop of a module during refueling comprised over 95 percent of total CDF. Consequently, the staff did not identify any structures, systems, and components (SSCs) that needed to be included under RTNSS. In contrast, in the US460 design, incomplete emergency core cooling system (ECCS) actuation dominates the core damage risk profile. Design changes in the US460 design result in increased ECCS demand, which increases the potential for incomplete ECCS actuation thereby resulting in core damage. Loss of AC power for longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> results in ECCS actuation.
The staff uses guidance contained in SRP 19.3, Revision 0, dated June 2014, Regulatory Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors, (ML14035A149) to conduct its review of an applicants RTNSS evaluation. NuScale uses this same guidance per Section 19.3 of the SDAA FSAR, which states, [t]he RTNSS scope, process, and criteria are consistent with the guidance of NUREG-0800 Section 19.3. In accordance with SRP 19.3, Revision 0 (page 19.3-6), staff responsible for the review of the applicants PRA verifies that the applicant has determined those nonsafety-related SSCs, if any, used to prevent the occurrence of initiating events and, based on their importance to risk as determined from the PRA, has included these nonsafety-related SSCs, as appropriate, in the scope of RTNSS.
SDAA FSAR Section 19.3.2.3 states, No nonsafety-related SSC are credited to meet NRC safety goals, to reduce the occurrence of initiating events, or to compensate for the uncertainties regarding passive systems in the PRA and in the modeling of severe accident phenomenology. Therefore, no nonsafety-related SSC meet the RTNSS C criteria.
Based on the staffs review of the SDAA FSAR and multiple documents available during the regulatory audit, the staff cannot verify that NuScale completely and satisfactorily addressed the following screening criteria, as stated on page 19.3-10 of SRP 19.3, Revision 0, in general, and for the two backup diesel generators (BDGs), in particular:
- 1. Does the calculation of the initiating event frequency consider the nonsafety-related SSCs?
- 2. Does the unavailability of the nonsafety-related SSCs significantly affect the calculation of the initiating event frequency?
- 3. Does the initiating event significantly affect the CDF and LRF (i.e., contribute to more than 10 percent of the at power or shutdown internal events CDF as stated in the footnote on page 19.3-10 of SRP 19.3, Revision 0)?
3 Based on its review of the event trees submitted in Chapter 19 of the SDAA FSAR, the BDGs are the only SSCs that completely avoid the need for ECCS actuation in the US460 design. The staffs review of the internal events CDF core damage sequences from the NuScale PRA, as reported in FSAR Table 19.1-17: Dominant Core Damage Sequences (Full Power, Internal Events, Single Module), identified that over 25 percent of the internal events CDF caused by losses of offsite power is mitigated by the two BDGs without the need to initiate ECCS.
Therefore, successful operation of the BDGs directly impacts the SDAA frequency of station blackout (SBO) and ECCS actuation for each NuScale Power Module (NPM). 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. Additionally, based on material available during the regulatory audit, the failure probabilities of the BDGs in the PRA are based on data for SBO DGs for operating plants. While the SBO DGs are scoped into the SBO rule for operating reactors, currently there is no regulatory program to ensure that the reliability and availability of the BDGs for the US460 design will be consistent with what is assumed in the PRA. Since loss of offsite power events are multi-unit events, the reliability of the BDGs impacts the defense-in-depth of the design.
From the staffs review of the material available during the regulatory audit, the staff also noted that the initiating event frequency for Loss of Offsite Power (LOOP) in the internal events PRA includes extratropical straight winds, F0 and F1 tornadoes, and Category 1 and 2 hurricanes.
The staffs review of material available during the regulatory audit identified that the two BDGs are Seismic Category III, which appear to be qualified for F0 and F1 tornadoes, and Category 1 and 2 hurricanes. These details are not included in the FSAR.
Information Requested:
To support the staffs finding against 10 CFR 52.137(a)(25) on the SDAAs conformance with the Commissions direction on RTNSS, NuScale is requested to:
Demonstrate how the RTNSS criteria related to the initiating event frequency stated on page 19.3-10 of SRP 19.3, Revision 0, dated 2014, are applied for the SDAA. Provide the results of NuScales evaluation against these criteria.
- 1. Since the successful operation of the two BDGs substantially decreases both the SBO and ECCS actuation frequency for all six NPMs in the US460 design and consequently, over 25 percent contribution to internal events CDF for a single NPM due to the potential for partial ECCS actuation, provide FSAR markups that (i) document the assessment of the two BDGs against the criteria on page 19.3-10 of SRP 19.3, Revision 0, dated 2014, (ii) identify the two BDGs as risk significant and nonsafety-related SSCs, (iii) identify the augmented quality and design details for the two BDGs, and (iv) identify the associated Inspections, Tests, Analysis, and Acceptance Criteria, or provide a means to ensure that the reliability of the two BDGs will be achieved in operation (e.g., a COL item to scope the BDGs within 10 CFR 50.65).
- 2. Provide FSAR markups for Chapter 19 that describe that the Seismic Category III building housing of the two BDGs is qualified to withstand F0 and F1 tornadoes, and Category 1 and 2 hurricanes consistent with the discussion in the PRA high winds notebook.