Licensing Guidance Frameworks

Objective / Purpose

Guidance is focused on surveying and developing sufficient general design criteria such that an advanced reactor design complies with licensing requirements developed by the Nuclear Regulatory Commission (NRC), or another regulatory body such as the Canadian Nuclear Safety Commission (CNSC) or the International Atomic Energy Agency (IAEA). It should also be mentioned that while the IAEA provides useful information regarding the licensing of advanced reactors across different countries, unlike the NRC the IAEA does not produce regulations and does not have any formal research groups to review its findings.

Scope

The NRC has developed several Reg Guides that apply to the adaptation of the Code of Federal Regulations (CFR) 10 Part 50 and 52 for the licensing of non-LWR (not water cooled) designs. Additionally, 10 CFR part 53 is currently under development for how it applies to advanced reactor licensing. Please visit the NRC Regulations Title 10 homepage for the full directory of title 10 regulations.

Guidance

New Research

In EPRI 3002032230 [71] risk-informed performance-based (RIPB) principles are applied to the structural design of advanced reactors using non-nuclear codes and standards (C&S) and supplemented by special treatments as needed.

• Example design cases in EPRI 3002032230 demonstrate key steps to selecting non-nuclear C&S provisions and special treatments, designing according to those selected provisions, and then evaluating the design to verify that it meets the reliability and capability targets. The design cases use a shear wall structure and show that non-nuclear C&S (for example, American Society of Civil Engineers/Structural Engineering Institute 7-22, American Concrete Institute 318-19), supplemented with special treatments as necessary, can achieve the necessary performance targets for safety-significant structures.
• The design example also includes a comparison between selected nuclear and non-nuclear C&S provisions to assess key differences for their impact on reliability and capability. A comparison of design criteria, provision by provision, reveals gaps and opportunities for flexibility in design criteria and can facilitate identification of helpful special treatments for structural design.
• Compatibility among C&S provisions, regulatory requirements, combined hazard effects, and quality assurance expectations are additional considerations for RIPB structural design.

Implementation of ASME Section XI Reliability and Integrity Management has a particular benefit for advanced reactor designs as it provides a tool to manage the uncertainty associated with the use of novel materials and novel operating conditions in these designs, even with limited operational data. EPRI EPRI 3002032238 [73] provides initial guidance concerning how the PRA model for an advanced reactor can be prepared and utilized to meet the requirements of a RIM Program.

The Licensing Modernization Project (LMP) framework detailed in the Nuclear Energy Institute (NEI) guidance document NEI 18-04 [8] is currently being used by some AR developers as the basis for implementing RIPB design. Despite the substantial level of detail available for the LMP framework in NEI 18-04 and supporting references, AR developers encounter challenges and desire additional technical guidance in several critical areas—particularly in the implementation of RIPB design to external hazards (RIPB-EH). To support and accelerate adoption of RIPB design methods, EPRI 3002029295 [74] evaluates ongoing industry activities on RIPB-EH, provides practical recommendations for its application, and identifies implementation challenges and research needs.

Historical

In the US, NRC Reg Guide 1.233 [9] provides guidance to applicants who are seeking licenses to operate advanced non-light water reactor (not water cooled) nuclear power plants and Reg Guide 1.232 [12] provides guidance on the licensing of non-LWRs from Appendix A of Title 10 Part 50 CFR. Although 10 CFR 50 Appendix A provides some guidance in establishing principal design criteria (PDC) for non-LWR (not water cooled) designs, it is the applicant’s responsibility to develop the PDC for its facility based on its design, using generic design criteria, non-LWR (not water cooled) design criteria, or other design criteria as the foundation. Similar to Reg Guide 1.232, INL/EXT-14-31179 [21] describes design guidance related to advanced non-light water reactors (not water cooled).

In Canada the CNSC framework for risk-informed decision-making is described in REGDOC-3.5.3 [18]. In 2019, the CNSC and the U.S. NRC signed a memorandum of cooperation (MOC) [19] to increase collaboration on technical reviews of advanced reactor and small modular reactor technologies.

Outside of North America, the IAEA has provided guidance in developing methodologies and safety requirements for AR designs. In particular, the current IAEA safety approach is based on four main pillars that are described and shown in Figure 1 of TECDOC-1570 [16]:

• Qualitative Safety Objectives
• Fundamental Safety Functions
• Defense in Depth
• Probabilistic Safety Assessment

TECDOC-1570 also proposes a new safety approach for advanced reactors based on the review of existing pillars to include any consideration of new technologies and incorporation of probabilistic considerations. In particular, the Safety Goals are identified in terms of consequences as a function of likelihood of occurrence.

In current practice, advanced reactor vendors develop a generic design phase PRA or a reference plant PRA. A Level 1 PRA or both Level 1 and Level 2 PRAs may be requested by the applicable regulatory authority for each new reactor project. A Level 3 PRA is typically not developed or requested by the regulatory authorities for new reactors. However, a Level 3 PRA could support the definition of emergency zones [25]. In the US, the Licensing Modernization Project (LMP) has developed a methodology that uses standard safety analyses like Process Hazards Analysis (PHA) and PRA to support various safety applications during the design and development process, including evaluating design alternatives, selecting Licensing Basis Events (LBEs), classifying structures, systems, and components (SSCs), and assessing the adequacy of defense-in-depth.