ML24297A138
| ML24297A138 | |
| Person / Time | |
|---|---|
| Site: | 99902100 |
| Issue date: | 10/22/2024 |
| From: | Stephanie Devlin-Gill Office of Nuclear Reactor Regulation |
| To: | TerraPower |
| References | |
| EPID L-2024-TOP-0005, CAC 000431 | |
| Download: ML24297A138 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION TERRAPOWER, LLC - AUDIT PLAN FOR TOPICAL REPORT: REACTOR SEISMIC ISOLATION SYSTEM QUALIFICATION, REVISION 0 (CAC/EPID NO. 000431/L-2024-TOP-0005)
Applicant:
TerraPower, LLC Applicant Address:
15800 Northup Way Bellevue, WA 98008 Plant Name:
Natrium Project No.:
99902100
Background:
By letter dated March 8, 2024, TerraPower, LLC (TerraPower) submitted topical report (TR)
NAT-8922, Reactor Seismic Isolation System Qualification Topical Report, Revision 0, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24068A212) to the U.S. Nuclear Regulatory Commission (NRC) staff. The TR describes the methodology and requirements necessary to establish the design criteria and qualification of the Natrium reactors seismic isolation system (SIS). On April 18, 2024, the NRC staff found that the TR contains sufficient technical information to enable the NRC staff to conduct a detailed technical review (ML24101A195).
Purpose:
The purpose of the audit is for the NRC staff to gain further understanding of TerraPowers SIS methodology and requirements. A secondary purpose of the audit is to identify any information that will require docketing to support the NRC staffs safety evaluation of the TR. Therefore, the NRC staff is requesting access to the TerraPower documents associated with the reactor SIS as discussed in the TR.
Regulatory Audit Basis:
The basis for the audit includes:
Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a)(3)(i) requires that facilities describe the Principal Design Criteria (PDC) in its preliminary safety analysis report supporting a construction permit application.
o Natrium PDC 2, as described in TR NATD-LIC-RPRT-002, Principal Design Criteria for the Natrium Advanced Reactor, Revision 1 (ML24101A362) relates to design bases for protection against natural phenomena. PDC 2 requires that safety-significant structures, systems, and components shall be designed to withstand the
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION effects or natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.
Regulation 10 CFR 100.23(c), as it relates to the investigation of seismic factors that may affect the proposed nuclear power plant.
RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion, Revision 0 (ML070310619), provides methods for defining site-specific, performance-based earthquake ground motion that satisfy the requirements of 10 CFR 100.23, Geologic and seismic siting criteria, and lead to the establishment of the safe shutdown earthquake ground motion (SSE).
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, Revision 0 (ML20091L698), 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. RG 1.233 endorsed Nuclear Energy Institute (NEI) 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1 (ML19241A472).
Regulation 10 CFR Part 50, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants is applicable to applications for a construction permit or operating license pursuant to 10 CFR Part 50 on or after January 10, 1997. Appendix S requires that for SSE ground motions, certain SSCs will remain functional and within applicable stress, strain, and deformation limits. The required safety functions of these SSCs must be assured during and after the vibratory ground motion through design, testing, or qualification methods. The evaluation must consider soil-structure-interaction effects and the expected duration of the vibratory motion. If the operating basis earthquake (OBE) is set at one-third or less of the SSE, an explicit analysis or design is not required. If the OBE is set at a value greater than one-third of the SSE, an analysis and design must be performed to demonstrate that when subjected to the effects of the OBE in combination with normal operating loads, all SSCs of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public must remain functional and within applicable stress, strain, and deformation limits.
Regulatory Audit Scope and Methodology:
This audit will focus on information provided by TerraPower in its online reference portal and during virtual meetings. This audit will provide the NRC staff with information related to the review of the SIS.
The regulatory audit will follow the guidance in the Office of Nuclear Reactor Regulation, Office Instruction LIC-111 Regulatory Audits, Revision 1 (ML19226A274), and focus on information provided by TerraPower in the online reference portal.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Information and Other Material Necessary for the Regulatory Audit:
The NRC staff requests TerraPower to make available the information and/or subject matter experts necessary to respond to the audit inquiries included below. These inquiries pertain to technical considerations presented in NUREG/CR-7253, Technical Considerations for Seismic Isolation of Nuclear Facilities (ML19050A422). Specific audit information needs and questions include but are not limited to the following:
- 1. Provide an overview of the seismic isolation approach of the Reactor Enclosure System (RES) using the spring and damper system, including the following:
(a) Configuration and arrangement of the isolation system, including the number of devices, anchorage to the Reactor Building (RXB), and other components.
(b) Preliminary plans for the isolator design, including optimization of the frequencies, balancing of horizontal and vertical response to minimize the rotational response, and determination of the initial properties of the spring and dampers.
(c) Approach to determining the SIS performance envelope, including ((
)) that will be used for design and testing of the isolation spring units (ISU) and isolation damper units (IDU) considering guidance in provided in NUREG/CR-7253 and American Society of Civil Engineers (ASCE) 43-19.
- 2. TR references [12] and [27] are not publicly available documents. The NRC staff requests that these be made available in the online reference portal.
- 3. NUREG/CR-7253, section 3.3, Types of Seismic Isolators Used for Base Isolation, provides a list of five qualifications tasks that should be accomplished before a new type of bearing is used to isolate [a nuclear power plant] in the US. Qualification task number 2 is Development of verified and validated numerical models capable of predicting the results of the dynamic testing of the prototype isolators. The TR provides a list of several structural analysis codes that implement the analytical models for the spring and viscous damper. Provide additional information on how the numerical models will be validated using the results of the prototype tests and what verification will be performed.
- 4. NUREG/CR-7253, section 3.3, qualification task number 5 is Verification and validation of numerical models of numerical tools to predict seismic response of the isolated systems. The TR discusses the computer modeling codes that can be used in the dynamic analysis. Confirm which software will be used for the prediction of the seismic response of the SIS and provide plans for verification and validation of the software.
- 5. NUREG/CR-7253 discusses three methods of analysis for seismically isolated structures. TR section 7.5, Reactor Seismic Isolation System Analysis, states the use of ((
)) approach to develop design parameters for the isolators and isolated SSCs, but does not provide sufficient detail on how this approach will be implemented. For example, it is unclear how the seismic demands are developed for the
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION isolators and isolated SSCs. Provide additional details on what type of preliminary and final design analyses will be performed affecting the SIS and isolated SSCs.
- 6. Describe how the inputs at the base of the SIS will be developed including the consideration of uncertainties. For example, how is the guidance on peak broadening in RG 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Revision 1 (ML003739367) implemented?
- 7. Technical Area section 3.3 in TR Table 7-1, Commentary on NUREG/CR-7253, outlines the SIS system-level qualification requirement recommended by NUREG/CR-7253 and mandated by ASCE 4-16, Seismic Analysis of Safety-Related Nuclear Structures, Chapter 12: Seismically Isolated Structures and ASCE 43-19, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities Chapter 9: Seismically Isolated Structures.
(a) Under Item d, system level testing, publication [29] does not directly provide information on system level testing. If the intent is to demonstrate system level performance based on the test results obtained from the literature, please discuss how the tests obtained from the literature adequately address system level testing for the SIS configuration, the spring and damper design and frequency, and amplitude consideration of the isolation system for the RES.
(b) Will system-level testing for the SIS be performed in 3D such that effects of potential rocking motions are accounted for? If not, provide technical justification to address why such testing is not deemed necessary or feasible and provide assurance that the safety and reliability of the SIS are adequately ensured through other means.
- 8. Table 7-1, Technical Area section 8.2, Performance Matrix, Isolators and Isolation System, Foundation, Umbilical Lines, Stop commentary states, in part, ((
)). A similar statement can also be found in Table 7-1, Technical Area section 8.5, Performance Expectations for Beyond Design Basis Ground Motion Response Spectrum Shaking, commentary a).
The NRC staff understands that the use of a physical stop is to serve as a fail-safe system to avoid the cliff edge effect for beyond design basis loading. Additionally, an engineered stop or displacement restraint is required by ASCE 4-16, section 12.5.3, Stop, and ASCE 43-19, section 9.3.3, Stop. Clarify how meeting the RIPB performance goal eliminates the need for the installation of a stop including the rationale and criteria that will be used to make these decisions. For example, how will you determine what is the appropriate low annual exceedance frequency (AEF) hazard? Will the design and testing account for the displacement resulting from the low AEF hazard?
- 9. Table 7-1, Technical Area section 9.1, Additional considerations, commentary a),
states, in part, ((
)). Although it can be inferred that it is meant to consider the variability in the mechanical properties of the seismic isolators as
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION required by ASCE 4-16, section 12.5.1, General, it is unclear how the variability of the isolator mechanical properties, such as stiffness, damping coefficient, and damping ratios, will be established and considered in the SIS analysis. Explain the determination of the isolator mechanical property variability and how the design parameters will be varied in the analysis to account for the property variation.
- 10. TR section 7.7, Reactor Seismic Isolation System Qualification, identifies the reactor SIS as active mechanical equipment that falls under the qualification program of American Society of Mechanical Engineers, Qualification of Active Mechanical Equipment Used in Nuclear Facilities, (ASME QME-1). However, the reliability and integrity management program described in section 7.8, Reactor Seismic Isolation System Lifetime Management, addresses lifetime management of passive components.
Describe how the active function of the SIS will be managed and discuss if the periodic in-service testing will be conducted to verify the functional performance of the SIS.
- 11. Related to lifetime management of the SIS, TR section 7.4.4.12, Seismic Isolators Long-Term Performance, states that ((
))
(a) ((
))
(b) NUREG /CR-7253 section 3.3, qualification task number 3 provides guidance for the demonstration of mechanical properties over a range of temperatures. TR section 7.4.4.12 addressed the acceptance criteria for material testing; however, it is not clear the range of temperature which the isolators will be tested. Confirm the range of temperatures for testing of the isolators. If different from the 40 degrees Fahrenheit (°F) (4.44 degrees Celsius (°C)) to 80°F (26.67°C) provided in NUREG/CR-7253, explain the basis for this difference.
- 12. TR section 2, Assumptions Requiring Verification and Open Items, states that there are no open items that require future actions to verify and close. However, the SIS RIPB design process includes steps such as: 1) seismic probabilistic risk assessment (SPRA) feedback to update SIS seismic special treatments to meet risk objectives and 2) update the SIS design and qualification and iterate on SPRA feedback. Please clarify if an SPRA has been developed. If it has, please make it available for the audit. If it has not, clarify the above statement indicating that there are no open items that require future actions.
- 13. TR figure 7-1, Seismic Isolation System risk informed performance-based design process, outlines a proposed methodology with three main steps: 1) update seismic PRA (SPRA) using preliminary design and corresponding fragilities; 2) SPRA feedback to update SIS seismic special treatments to meet risk objectives; and 3) update SIS design and qualification and iterate on SPRA feedback. Please describe how the SIS will be modeled in the SPRA to support these steps. To ensure the feasibility of these steps, the NRC staff seeks clarification on the following areas:
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (a) How TerraPower defines the failure modes for the active components (ISUs) and the passive components (IDUs);
(b) How the failure probabilities for these units will be determined; (c) What the success criteria are (i.e., the minimum number of units that must function properly for success);
(d) What reliability assurance activities are in place to prevent unacceptable degradation of the units; (e) How the fragility analysis of these units will be performed; and (f) What the consequences would be in the event of ISU or IDU failure.
- 14. TR section 6, Natrium Reactor Seismic Isolation System, states that the primary system to be supported by the SIS is the RES and provides a high-level description of the RES umbilical lines. TR section 7.4.6.3, Umbilical Lines Crossing the Isolation Interface, provides a requirement for safety-significant umbilical lines to accommodate the maximum displacement of the SIS, but no information on the design approach was provided. The NRC staff is seeking additional information on the seismic impacts on the equipment attached to the RES and the design aspects used to prevent the release for these safety-significant connections, including:
(a) Discuss if breaks or disconnections to the equipment could result due to different vibrations and movements; (b) Describe how potential failures of the lines that cross the SIS boundary are treated in the safety analysis, and (c) Provide additional information on the design approach for the umbilical lines, for both mechanical and electrical systems.
- 15. The relationship between EPZ/Cliff Edge and DID Only to licensing basis events in TR figure 7-4, Licensing Modernization Project Event Types by Frequency of Event, does not appear to be consistent with the Licensing Modernization Project methodology.
Please discuss this figure.
Team Assignments:
Meg Audrain Materials Engineer Nilesh Chokshi Civil/Structural Engineer, contractor to NRC, Center for Nuclear Waste Regulatory Analyses (CNWRA)
Biswajit Dasgupta Civil/Structural Engineer, contractor to NRC, CNWRA Stephanie Devlin-Gill Senior Project Manager, Audit Manager Bruce Lin Mechanical Engineer Hanh Phan Senior Reliability and Risk Analyst
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Tracy Radel Nuclear Engineer, Audit Lead John Stamatakos Structural Geologist and Geophysicist, contractor to the NRC CNWRA Logistics:
Entrance Meeting November 7, 2024 Exit Meeting January 31, 2025 Audit meetings will take place in a virtual format, using Microsoft Teams or another similar platform. Audit meetings will be scheduled on an as-needed basis after the entrance meeting and once the NRC staff has had the opportunity to review any documents placed in the online reference portal. The audit will begin on November 7, 2024, and continue for approximately 2 months, with activities occurring intermittently during the audit period. The audit period may be reduced or extended, depending on the progress made by the NRC staff and TerraPower in addressing the audit questions.
An online reference portal, established by TerraPower, would allow the NRC staff to access technical information. Use of the online reference portal is acceptable provided that TerraPower establishes measures to limit access to specific NRC staff (e.g., based on NRC email addresses or the use of passwords which will only be assigned to the NRC staff or contractors directly involved in the audit on a need-to-know basis) and to make the documents view-only (i.e.,
prevent the NRC staff from saving, copying, downloading, or printing any documents). The conditions associated with the online reference portal must be maintained throughout the audit process. The NRC staff who should initially be granted access to the online reference portal are listed in the Team Assignments section above. The NRC audit manager will provide TerraPower with the names of additional NRC staff who should be granted access.
Special Requests:
None Deliverables:
At the completion of the audit, the audit team will issue an audit summary within 90 days after the exit meeting but will strive for a shorter duration. The audit summary will be declared and entered as an official agency record in ADAMS and be made available for public viewing.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION If you have questions regarding this audit, please contact me at 301-415-5301 or via email at Stephanie.Devlin-Gill@nrc.gov.
Date: October 22, 2024 Sincerely,
/RA/
Stephanie Devlin-Gill, Senior Project Manager Advanced Reactors Licensing Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Project No.: 99902100 cc: TerraPower Natrium via GovDelivery
Package: ML24297A136 Email: ML24297A140 Public: ML24297A138 Non-Public: ML24297A139 NRR-106 OFFICE NRR/DANU/UTB2 NRR/DANU/UAL1:PM NRR/DANU/UTB2:BC NAME TRadel SDevlin-Gill CdeMessieres DATE 10/02/2024 10/02/2024 10/10/2024 OFFICE NRR/DANU/UAL2:LA NRR/DANU/UAL1:BC NRR/DANU/UAL1:PM NAME CSmith JBorromeo SDevlin-Gill DATE 10/22/2024 10/15/2024 10/22/2024