ML24051A030
| ML24051A030 | |
| Person / Time | |
|---|---|
| Site: | 99902100 |
| Issue date: | 01/30/2024 |
| From: | Brusselmans R NRC/NRR/DANU/UAL1 |
| To: | George Wilson TerraPower |
| References | |
| EPID L-2023-TOP-0020 | |
| Download: ML24051A030 (1) | |
Text
Enclosure OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY AUDIT
SUMMARY
REPORT TERRAPOWER, LLC TOPICAL REPORT PRINCIPAL DESIGN CRITERIA FOR THE NATRIUM ADVANCED REACTOR, REVISION 0 PROJECT NO. 99902100
1.0 BACKGROUND
By letter dated January 24, 2023, TerraPower, LLC (TerraPower) submitted topical report (TR)
NATD-LIC-RPRT-002, Principal Design Criteria for the Natrium Advanced Reactor, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23024A281) to the U.S. Nuclear Regulatory Commission (NRC) staff. The NRC staff finalized its completeness determination of the TR on March 17,2023, and found that the TR contained sufficient information such that the NRC staff could begin its detailed technical review of the TR (ML23074A349).
TerraPower requested the NRC staffs review and approval of its proposed principal design criteria (PDC), as presented in the TR. These PDC would be used by applicants using TerraPowers Natrium reactor design as part of future licensing submittals. TerraPowers overall licensing approach for the Natrium reactor design follows the Licensing Modernization Project (LMP) methodology described in Nuclear Energy Institute (NEI) 18-04, Revision 1, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development (ML19241A472). Regulatory Guide (RG) 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light Water Reactors, Revision0 (ML20091L698) endorses the LMP methodology described in NEI 18-04. Additionally, PDC development was informed by RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, Revision 0 (ML17325A611).
2.0 AUDIT REGULATORY BASIS The basis for the audit included the following:
Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a)(3)(i) requires that each application for a construction permit include PDC for the facility as part of a description of the facilitys preliminary design information.
Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, contains minimum requirements for the PDC for water-cooled nuclear power plants. These general design criteria (GDC) are generally applicable to non-light-water reactors (non-LWRs) and are intended to provide guidance in establishing the PDC for non-LWRs.
RG 1.232 provides guidance on how the GDC may be adapted for non-LWR designs.
The RG also describes the NRCs proposed guidance for modifying and supplementing
2 the GDC to develop PDC that address two specific non-LWR design concepts: sodium-cooled fast reactors (SFRs) and modular high temperature gas-cooled reactors.
As noted above, RG 1.233 endorses the use of NEI 18-04 and provides guidance on using a technology-inclusive, risk-informed, and performance-based methodology to inform the licensing basis and content of applications for non-LWRs.
3.0 AUDIT PURPOSE AND OBJECTIVES The purpose of the audit was for the NRC staff to gain a more detailed understanding of the PDC proposed for the Natrium SFR. The audit was also performed to identify any information that will require docketing to support the NRC staffs development of the safety evaluation for the TR. The audit plan, including the initial set of audit questions, was issued on August 22, 2023 (ML23201A247).
4.0 SCOPE OF THE AUDIT AND AUDIT ACTIVITIES The audit was conducted from September 7, 2023, through October 26, 2023, in a virtual format and followed the guidance in the Office of Nuclear Reactor Regulation Office Instruction LIC-111, Regulatory Audits (ML19226A274). TerraPower responded to the audit plan questions over the course of multiple audit meetings. To support addressing NRC staff questions, TerraPower made the following documents available in its electronic Reading Room (eRR):
TP-LIC-LET-0066, NAT3418, Mechanistic Source Term and Radiological Consequences Method Development Update, TerraPower Presentation Slides in PDF form.
NATD-LIC-PRSNT-0017, Functional Containment.
TP-LIC-PRSNT-0007, [Specified acceptable system radionuclide release design limits (SARRDLs)], Functional Containment and Major Accident Update.
NRC Technical Letter Report (TLR)RES/DE/CIB-2019-01, Advanced Non-Light-Water Reactors Materials and Operational Experience, (ML18353B121).
NRC TLR-RES/DE/CIB-CMB-2021-07, Corrosion in Sodium Fast Reactors, (ML21116A231).
International Atomic Energy Agency Report, Sodium Fires at Fast Reactors.
NAT-LIC-PRSNT-0043, Reactivity Control.
In addition, TerraPower provided written summaries of its response to some of the NRC questions in the eRR.
Members of the audit team included the NRC staff listed below.
Reed Anzalone Senior Nuclear Engineer, Audit Lead Mallecia Sutton Senior Project Manager, Lead Natrium Project Manager (PM)
Stephanie Devlin-Gill Senior Project Manager, Audit PM Roel Brusselmans Project Manager, Audit PM Mohsen Khatib-Rahbar Consultant (Energy Research, Inc.)
Imtiaz Madni Consultant (Energy Research, Inc.)
The primary participants from TerraPower for this audit included Patrick Donnelly, Christopher Forrest, Kevin Gagne, Jeff Robertson, Brian Johnson, and Donald Lewis.
3 On October 26, 2023, the NRC staff held an audit exit meeting with TerraPower and summarized the audit purpose, activities, and high-level results. The NRC staff did not acquire any documents during the audit.
5.0
SUMMARY
OF OBSERVATIONS During the audit, TerraPower discussed its intended approaches to show compliance with regulatory requirements and the proposed PDC, including the use of the NEI 18-04 process, insights from the Natrium probabilistic risk assessment (PRA), and the current state of the Natrium design. A summary of NRC audit observations is provided below.
- 1. Chemical Stability and Potential for Corrosion of Liquid Sodium as a Coolant The NRC staff noted that the proposed TerraPower PDC do not explicitly address the potential adverse effects from interactions between the coolant and fuel, coolant and the coolant boundary, and coolant and the atmosphere; such considerations could merit their own PDC, similar to that specified in SFR Design Criteria (SFR-DC) 74 in RG 1.232. While TerraPower stated that the coolant is chemically stable, it has not demonstrated this feature at this stage of design and review. It is recognized that TerraPower is considering the available information related to the operational experience with SFRs (e.g., as discussed in NRC TLR-RES/DE/CIB-CMB-2021-07). The literature indicates that the corrosion of structural material in liquid sodium is strongly affected by the presence of oxygen and carbon impurities in the sodium, and it is strongly affected by the structural material composition, in some cases requiring additional testing and assessment.
Alternatively, there does not appear to be any consequential chemical interactions between the metallic fuel and sodium. However, there are known chemical interactions between the fuel and cladding (referred to as fuel cladding chemical interaction or FCCI). The effect of FCCI on the fuels ability to retain radionuclides is addressed by PDC through the SARRDLs, as discussed under item 4 below. It is recognized that as the Natrium design matures and the structural material selection is complete, the verification of interactions between the coolant and fuel and various structures and cladding will need to be considered to ensure the proposed PDC represent an adequate set of criteria for Natrium.
- 2. Functional Containment and Integrity of the Reactor Coolant System The Natrium primary coolant system is encompassed by the reactor enclosure system that includes the reactor guard vessel and the inerted annulus region. The reactor vessel and vessel head are the safety-related reactor coolant boundary components that will be designed to meet the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Class A requirements. The guard vessel is considered a safety-related component, but its safety function is associated with its effectiveness to transfer heat to the safety-related heat removal system (i.e., the reactor air cooling system). The guard vessels function to contain the sodium pool in the event of a reactor vessel leak is a defense-in-depth (DID) function, and TerraPower indicated that appropriate graded quality requirements would be applied, as necessary, to achieve the DID function. As such, TerraPower currently plans to design the guard vessel in accordance with ASME BPV Code Section VIII, Rules for Construction of Pressure Vessels.
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- 3. Guard Pipes for the Intermediate Coolant System and Implications for Sodium Fires The Natrium intermediate loop guard piping within the reactor vessel head access area is currently not intended to be inerted. TerraPower acknowledged that there is the potential for increased pressures and temperatures within the guard pipe annulus; therefore, it plans to address these with appropriate design control measures to prevent any adverse conditions.
In addition, for areas expected to require inspections (e.g., primary heat transport portion of the pipes that connect to the reactor head), a clamshell design is planned to facilitate removal and inspection.
With respect to Natrium PDC 73 and 74 for the intermediate sodium loop piping outside of the head access area, TerraPower plans to assess the sodium fire risk considering appropriate sodium leak containment design options (i.e., non-inerted guard piping removable or in-place, pipe jackets with catch pans, and suppression decks). Furthermore, TerraPower plans to design the non-inerted piping to withstand the calculated pressure loads from any sodium-air reaction within the annulus region. TerraPower stated during the audit that the primary purpose of pipe jackets is to ensure that the velocity of sodium from the intermediate coolant system is dissipated such that the sodium stream becomes a controlled gravity-drained flow instead of a jet spray. Catch pans will also be provided to prevent the interaction of sodium with water inside the concrete surroundings. The suppression decks over the catch pans are intended to minimize air circulation and the sodium-air reaction rate. These are in addition to planned systems for sodium leak, fire, and smoke detection to support operator response to minimize the volume of any sodium leakage.
The NRC staff also noted that the reactor building and reactor auxiliary building (RAB) are physically separated by some distance, and each building will be designed with a three-hour fire barrier rating. Also, most of the intermediate loop is located inside the RAB, and it will be designed (e.g., placement of catch pans) to retain most of any sodium that drains into the building, thus, preventing any contact of sodium with the concrete floor (i.e., to prevent reaction of sodium with water inside the concrete that produces hydrogen).TerraPower envisions that the various design considerations, leakage, fire, and smoke detection system, implementation of the maintenance requirements of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants and the emergency preparedness programs defined in 10 CFR 50 Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, assure safety of the Natrium plant.
- 4. Relationship between Fuel Design Limits and Limits on Release of Radionuclides The fuel design criteria are surrogates for the SARRDLs, with fuel damage criteria set to prescribe the most limiting set of fuel conditions to consider as the initiation point for safety analysis, and the failure criteria are used to identify failures that would need to be considered as radionuclide releases when assessing SARRDLs. In practice, design basis accidents are analyzed by TerraPower assuming the most limiting initial conditions and use a time-at-temperature criterion that is well below the actual fuel limits (i.e., there is sufficient margin to assume no fuel failure if the transient stays below the criterion and therefore further analysis is not performed). If the criterion is exceeded, then further analysis is performed to determine if fuel failure is predicted and, if so, this is used to inform the radionuclide release assessment.
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- 5. Reactivity Control System TerraPower provided information regarding the reliability of the reactivity control system as assessed in the Natrium PRA, indicating that all control rods share common cause failure (CCF) of binding and CCF failure of control. TerraPower stated that Defense Line analysis is used in part to demonstrate sufficient DID plant capability, which does not directly impact PDC 26 functions, and precedes the PRA, but updated with the PRA results. The defense line evaluation prepared by TerraPower discusses the ability to overcome Decoupling Failure that includes CCF to release the coupling and minor absorber to duct binding. It further states that the PRA results currently show that the frequency and consequence target goals are met with identical primary and secondary absorber bundle designs. It highlights the main objectives of the secondary control rod assembly design to provide DID adequacy with respect to CCF of absorber bundle to duct binding.
The NRC staff notes that a detailed assessment of the design was beyond the scope of the audit and, as such, the NRC staff did not perform a full review of the safety classification of different aspects of the reactivity control system.
TerraPower plans to perform SCRAM analyses and testing that include assessment of the impact of thermal-hydraulic, mechanical, and seismic effects (e.g., duct deformation) on SCRAM time. This is an important consideration to demonstrate subcriticality using the proposed reactivity control system design under normal and adverse conditions.
The potential for any dependencies of the primary and secondary reactivity control systems on the electrical or other systems (e.g., instrumentation and control systems) will be considered as part of future evaluations to demonstrate that these other systems will meet the electrical and other related design requirements.
- 6. Transient and Accident Categories and Relationship to the Licensing Basis Event Categories of NEI 18-04 TerraPower provided information stating that anticipated operational occurrences, as used in the proposed Natrium PDC, are consistent with the equivalent licensing basis event (LBE) category from NEI 18-04. TerraPower also stated that the various types of accidents discussed, including accidents, postulated accidents, and accident conditions, all refer to events in the design basis accident (DBA) category from NEI 18-04. The DBAs are derived from design basis events, as discussed in NEI 18-04.
This information clarified the intended scope of the proposed PDC for the NRC staff, since the DBAs are used to establish the requirements for safety-related structures, systems, and components (SSCs), as discussed in RG 1.233 and NEI 18-04. However, Appendix A to 10 CFR 50 states, in part, that the PDC establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety The use of the term important to safety in the regulations establishes a broader scope of SSCs to which the PDC must apply, which could include certain SSCs that would be categorized as non-safety-related with special treatment (NSRST) under the NEI 18-04 process. Under NEI 18-04, the requirements for these NSRST SSCs may be set by LBEs other than DBAs (i.e., beyond design basis events). As such, the PDC provided may not represent a full set of PDC for all SSCs that are important to safety and additional criteria may be needed for some NSRST SSCs. This is discussed in
6 additional detail in Staff Position C.6 of Draft RG DG-1404, Guidance for a Technology-Inclusive Content-of-Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors (ML23194A194). The NRC staff informed TerraPower that it plans to add a Limitation and Condition to the TR safety evaluation regarding this topic.
There were no deviations from the audit plan.
6.0 REQUESTS FOR ADDITIONAL INFORMATION RESULTING FROM AUDIT As a result of the audit, the NRC staff did not identify any specific items as requests for additional information related to this TR, nor did TerraPower identify a need to provide a voluntary revision or supplement to the TR. However, the NRC staff will continue to have interactions with TerraPower as part of the TR review process.
7.0 OPEN ITEMS AND PROPOSED CLOSURE PATHS There are no open items resulting from this audit.