ML24262A235
| ML24262A235 | |
| Person / Time | |
|---|---|
| Site: | 99902098 |
| Issue date: | 10/04/2024 |
| From: | William Williams Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML24262A233 | List: |
| References | |
| EPID L?2024?LRO?0029 | |
| Download: ML24262A235 (1) | |
Text
Enclosure U.S. NUCLEAR REGULATORY COMMISSION STAFF FEEDBACK REGARDING GENERAL ATOMICS ELECTROMAGNETIC SYSTEMS WHITE PAPER: FAST MODULAR REACTOR SAFETY CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS (EPID L2024LRO0029)
SPONSOR INFORMATION Sponsor:
General Atomics Electromagnetic Systems Sponsor Address:
16530 Via Esprillo San Diego, CA 92127 Project No.:
99902098 DOCUMENT INFORMATION Submittal Date: June 10, 2024 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML24162A320 Purpose of the White Paper: The purpose of this white paper (WP), Fast Modular Reactor Safety Classification of Structures, Systems, and Components, is to provide an overview of the regulations and available guidance that General Atomics Electromagnetic Systems (GAEMS) will be using to develop their approach to the process of safety classification of structures, systems, and components (SSCs) applicable to their helium-cooled fast modular reactor (FMR) design. The WP provides an interpretation, or summary, of the technology-inclusive, risk-informed, and performance-based approach found in Nuclear Energy Institute (NEI) 1804, Revision 1, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, (ML19241A472), which documents the result of the Licensing Modernization Project (LMP), as well as regulations relevant to SSCs. The WP also briefly discusses GAEMSs licensing basis development process and how probabilistic risk assessment (PRA) techniques will be used in the safety classification process. The WP then provides GA-EMSs preliminary process for selecting and classifying SSCs, a discussion of baseline required and safety significant functions, and an initial list of safety-related and non-safety-related with special treatment (NSRST) SSCs and functions.
Action Requested: GAEMS requested that the U.S. Nuclear Regulatory Commission (NRC) staff review the WP with no specific feedback requested.
FEEDBACK The NRC feedback on this WP is preliminary and subject to change. The feedback does not constitute any regulatory findings on any specific licensing matter and are not official NRC positions. The lack of feedback regarding a certain aspect of the WP should not be interpreted as NRC agreement with GAEMSs position. General feedback, including comments on SSCs, SSC classification, and the overarching process outlined by GAEMS, is provided below.
Regulatory Requirements and Applicable Guidance:
Assuming licensing under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and approvals for Nuclear Power Plants, the WP identifies appropriate regulatory bases for the determination and classification of SSCs. The regulations cited included:
- 1. 10 CFR 50.34, Contents of applications; technical information
- 2. 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors
- 3. 10 CFR Part 50 appendix A, General Design Criteria for Nuclear Power Plants
- 4. 10 CFR Part 50 appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
- 5. 10 CFR 52.47, Contents of applications; technical information
- 6. 10 CFR 52.137, Contents of applications; technical information Various additional regulations that the NRC staff notes may be used to inform SSC classification include the following:
10 CFR 50.2, Definitions, provides a link between other licensing areas and SSCs, such as how SSCs are used to determine design bases as well as the definition of safety-related SSCs.
10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, requires licensees under 10 CFR Part 50 and 10 CFR Part 52 to monitor the performance or condition of certain SSCs to ensure they can perform their intended functions. While 10 CFR 50.65 does not provide a process for the classification of SSCs, the scope of SSCs to which 10 CFR 50.65 must be applied depends on their safety significance.
Should a combined license application be of interest to GAEMS, 10 CFR 52.79, Contents of applications; technical information in final safety analysis report, describes requirements for the safety analysis of SSCs.
The WP also cites various Office of the Secretary (SECY) papers, as well as Regulatory Guide (RG) 1.233, Revision 0, 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, (ML20091L698), which endorses NEI 18-04. The NRC staff notes that these documents provide an acceptable approach for licensing basis event (LBE) selection and SSC classification.
However, the WP also refers to RG 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, (ML061090627). Either RG 1.233 or RG 1.201 constitutes an acceptable approach to determine the safety classification of SSCs, but the NRC staff expects that an applicant would choose one or the other. If GAEMS intends to use the risk-informed safety categorization of SSCs described in 10 CFR 50.69, then RG 1.201 provides the appropriate reference; if GAEMS intends to follow the LMP framework, NEI 18-04 and RG 1.233 are appropriate.
The remainder of the WP appears, without explicit statement, to be following the LMP approach because the safety classifications (i.e., safety-related, NSRST, and non-safety-related with no special treatment (NST)) are consistent with the LMP. Therefore, the NRC staff assumed LMP to be the planned approach and tailored its comments accordingly.
Licensing Basis Development Process The WP recounts the frequency-consequence (FC) evaluation criteria found in NEI 18-04 with minimal interpretation. The NRC staff notes that the relevant FC ranges and definitions for anticipated operational occurrence, design-basis event, and beyond-design-basis event are consistent with NEI 18-04. However, the WP states, [t]he F-C curve provides an acceptable limit in terms of the frequency of potential accidents and their associated consequences.
RG 1.233 states, [t]he staff emphasizes the cautions in NEI 18-04 that the F-C target figure does not depict acceptance criteria or actual regulatory limits. Therefore, the FC targets may be used to inform and support processes such as SSC classification and defense in depth, but GAEMS should not interpret them as NRC acceptable limits.
Process for Selecting and Classifying SSCs The WP recounts the possible safety classifications as found in NEI 18-04 verbatim and adequately paraphrases the seven tasks for the SSC safety classification process as outlined in NEI 18-04 with no notable discrepancies aside from level of detail. The WP states that GA-EMS also submitted another WP, Fast Modular Reactor Selection of Licensing Basis Events (ML24011A221), to describe the LBE selection process; the NRC provided feedback for that WP on June 26, 2024 (ML24164A088). The NRC staff notes that NEI 18-04 provides a process for LBE selection that is intended to be applied iteratively along with SSC classification as part of design maturation (see NEI 18-04, figure 3-2, Process for Selecting and Evaluating Licensing Basis Events).
Required and Safety Significant Functions and SSC Classification The WP recounts the definition of required safety functions (RSFs) from NEI 18-04 and proposes RSFs for the FMR. The NRC staff notes that the RSFs presented are closer to the level of fundamental safety functions (FSFs) from NEI 18-04 (i.e., control heat generation, control heat removal, retain radionuclides) than the definition of RSFs. In the NEI 18-04 process, FSFs are used to develop the PRA. The PRA then models PRA safety functions (PSFs), some of which are RSFs. As such, RSFs should be at a very detailed and design-specific level, which is not found in this WP. The NRC staff understands that this WP only outlines the process GA-EMS intends to use and that detailed PSFs and RSFs need not necessarily be discussed at this time. Nevertheless, this observation underscores the value of GAEMS continuing the iterative PRA process of LBE selection for engagement with the NRC staff at a more detailed level.
SSC Classification The NRC staff is unable to provide substantial feedback on FMR safety-related SSCs and functions, NSRST SSCs and functions, or NST SSCs (which are not included in this WP) due to lack of detailed design information, PRA analysis, and appropriately defined RSFs. For example, it is unclear how GA-EMS determined control rods to be safety-related while their associated control rod drive mechanisms are considered NSRST. This is not consistent with how such SSCs would be evaluated for almost all known reactor designs; however, it is possible that the LMP process could result in such an outcome. Under LMP, the safety classification would depend upon reactor trip mechanisms, ability to reach safe shutdown conditions, and other RSFs. The plants capability must be evaluated through PSFs, event trees, LBE selection, system boundaries, failure probabilities, etc. Without these details, there is no basis for the NRC staff to interpret SSC classification. It is also important to note that, specifically with respect to the reactivity control systems, expectations regarding safety classification of relevant SSCs are built into FMR design criterion 26, as discussed in the NRC staff approved GA-EMS topical report 30599T00005, Fast Modular Reactor Principal Design Criteria, Revision 2-A (ML23215A266).
Principal Contributors: Walter Williams, NRR Reed Anzalone, NRR