ML25182A209
| ML25182A209 | |
| Person / Time | |
|---|---|
| Issue date: | 07/03/2025 |
| From: | Gheen A NRC/NRR/DANU/UARP |
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| References | |
| Download: ML25182A209 (1) | |
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May 1, 2025 Version 1
Technology Inclusive Management of Safety Case (TIMaSC)
White Paper Outline I.
Introduction A.
Purpose and Scope The White Paper is intended to develop a shared understanding of the key elements of guidance on management of the risk-informed, performance-based (RIPB) licensing basis. The paper will form the foundation for a subsequent guidance document that will be submitted to the Nuclear Regulatory Commission (NRC) for endorsement. The intended audience for the White Paper is the NRC, in order to obtain initial feedback, and advanced reactor vendors and potential operators, in order to promote industry alignment.
Developing the White Paper will involve an assessment of existing processes and programs for maintaining the RIPB licensing basis of an advanced reactor. Processes and programs for currently operating light water reactors (LWRs) were developed for their predominantly deterministic licensing bases and may require adjustment for application to advanced reactors with RIPB licensing bases. The paper should identify where there may be gaps that need to be addressed by additional guidance.
B.
Perspective on Technology Inclusive RIPB Regulation The NRC regulatory framework has evolved to include RIPB elements over the past 30 years. PRA has become a tool that is routinely used to support operational decision-making in the current fleet. PRA insights are being incorporated into the design of a new generation of advanced reactors. As discussed in the background section, the TIMaSC Project builds upon earlier guidance that established an acceptable RIPB licensing approach for advanced reactors. The probabilistic risk assessment (PRA) underlying this approach will be a focus area for the TIMaSC guidance.
C.
Background
The section will discuss existing industry and NRC guidance for licensing advanced reactors with an RIPB approach.
- 1.
Licensing Modernization Project (LMP): NEI 18-04 and NRC Regulatory Guide (RG) 1.233
- 2.
Technology Inclusive Content of Application Project (TICAP): NEI 21-07 and NRC RG 1.253
- 3.
NRC Advanced Reactor Content of Application Project (ARCAP)
- 4.
Technology Inclusive Risk Informed Change Evaluation (TIRICE): NEI 22-05
- 5.
Other past and ongoing risk-informed, performance-based initiatives for advanced reactors
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II.
Advanced Reactor RIPB Licensing Basis This chapter will discuss, in detail, the holistic approach to the advanced reactor licensing basis, including the treatment of defense-in-depth. It will introduce the Network Diagram (see Figure 1 for illustration) and address the relationship between the different elements. It will also discuss the interaction between deterministic elements and risk-informed elements in the licensing basis.
III.
Facility Changes to Which NEI 22-05 is Applicable Consistent with NEI 96-07 for the operating fleet, NEI 22-05 is not applied to facility changes that are controlled by more specific requirements and criteria established by regulation (see NEI 22-05, Section 4.1.1). An example is a change to the physical security plan, which is controlled under 10 CFR 50.54. Thus, this section applies to facility changes that would go through the screening and, if necessary, evaluation processes established in NEI 22-05.
This section will address specific measures to be taken for facility changes once the need for a license amendment is determined (either way). The section will discuss addressing impacts on definition, frequency, and consequences of LBEs; defense-in-depth evaluation; reliability and capability targets; special treatments; and other elements of the licensing basis.
IV.
Facility Changes to Which NEI 22-05 is Not Applicable The NEI 22-05 process would be used to determine the need for a license amendment for facility changes that are not controlled by more specific requirements and criteria established by regulation. Examples of such categories of facility changes are quality assurance and quality implementing activities, emergency planning, and physical and cyber security.
This section will address specific measures to be taken for facility changes addressed by more specific regulations. The section will discuss addressing impacts on definition, frequency, and consequences of LBEs; defense-in-depth evaluation; reliability and capability targets; special treatments; and other elements of the licensing basis.
V.
New Information NEI 22-05 discusses new information in Section 4.1.6, stating New Information is routinely acquired as a natural part of the design, operation, and maintenance of nuclear facilities. New information for new reactor designs is expected to be identified as a result of the compilation of the knowledge from operating experience, experiments, and testing.
LWRs have been operating in the United States for nearly 70 years. The LWR technology is mature. As a result, far less new information arises today than did during the initial period of LWR operation. The experience bases for advanced reactors expected to be licensed using a RIPB framework varies among the designs, but those experience bases will not be nearly as extensive as the LWR experience base. Accordingly, it is expected that new information will arise periodically for new designs, and it will need to be addressed within the context of the reactor licensing basis.
May 1, 2025 Version 3
New information by itself is not a facility change. NEI 22-05 does not provide guidance on addressing new information, except to the extent it has resulted in a facility change that is to be evaluated. This section of the white paper will address, at a high level, the processes to be followed for addressing different types of new information. These processes are expected to be very similar to the processes developed over the years for LWRs, although the implementation details may be different due to the RIBP licensing basis.
The biggest direct impact of new information is expected to be on the plant PRA, which is an integral part of the RIPB licensing approach as outlined in NEI 18-04. Accordingly, Chapter VI on PRA includes a subsection addressing new information as it relates to the PRA.
VI.
Probabilistic Risk Assessment (PRA)
PRAs for operating LWRs typically focus on the frequencies of events that could lead to core damage (i.e., Level I PRAs). The LWR licensing basis is largely deterministic and not affected by the plant PRA. For advanced reactors following NEI 18-04, the PRA plays an important role in establishing and maintaining the licensing basis. Advanced reactor PRAs model event consequences to a much greater extent than a Level 1 LWR PRA. It is expected that advanced reactor PRAs will conform to the guidance in the advanced non-LWR PRA Standard ASME/ANS RA-S-1.4-2021. That standard addresses many, but not all, issues associated with ensuring that the advanced non-LWR PRA is maintained and used in a manner consistent with the plant licensing basis.
This chapter of the White Paper will address the impact of changes on the PRA and the role of the PRA in maintaining the plant licensing basis. Topics to be covered are listed below.
A.
Evolution of Advanced Reactor PRAs B.
Role of PRA in Establishing the Licensing Basis under NEI 18-04 C.
Non-LWR PRA Standard and Supporting Peer Reviews D.
PRA Configuration Control, Updates and Change Management E.
New Information F.
Cautions and Limitations VII. Other Topics This section will discuss specific topic areas that are potentially related to the licensing basis of an advanced reactor following the LMP licensing approach, but for which no decision has been made to include it in the TIMaSC guidance document. These topics include areas for which the development of additional guidance is planned or underway, but not complete. In such cases, this White Paper will identify the planned or ongoing activities. These topics also include areas that may be appropriate for guidance for a licensee using the LMP methodology, but for which no guidance is under development or planned.
Examples of other topics are provided below.
Risk-informed Technical Specifications
May 1, 2025 Version 4
Reliability Assurance Program Reactor Oversight Process Fire Protection Risk Metrics Considerations Prior to Operation Adding Units to a Site VIII. Summary This chapter will summarize the information presented in the White Paper and present conclusions that are pertinent to the subsequent development of the TIMaSC guidance document.
May 1, 2025 Version 5
FIGURE 1 Network Diagram