ML24268A317

Draft White Paper on Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations - Enclosure 3

ML24268A317
Person / Time
Issue date:	09/27/2024
From:	Harvey J, William Kennedy NRC/NRR/DANU/UARP
To:
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Download: ML24268A317 (22)

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Preliminary White Paper - Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, Enclosure 3 (September 2024)

Preliminary White Paper - Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, Enclosure 3 (September 2024) to NRC Staff Prepared White Paper Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations September 2024 Draft - Released to Support ACRS Interaction THIS NRC STAFF WHITE PAPER HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS). THIS PAPER HAS NOT BEEN SUBJECT TO NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS. A PROSPECTIVE APPLICANT SHOULD NOT USE THE CONTENTS OF THIS PAPER OR RELY ON ITS CONTENTS IN PREPARING AN APPLICATION.

This enclosure includes various topics related to the licensing and deployment of nth-of-a-kind (NOAK) micro-reactors (and potentially applicable to other new reactors, as appropriate). Some of these are potentially generic topics raised by developers through formal pre-application engagement with the U.S. Nuclear Regulatory Commission (NRC) staff and in other interactions, such as the periodic Advanced Reactor Stakeholder Meetings organized by the NRC staff. The NRC staff will address design-specific issues on a case-by-case basis.

This enclosure includes the NRC staffs plans for addressing the topics. These include means for the NRC staff to address each topic under the existing regulatory framework and longer-term approaches. The NRC staff will engage the Commission on any future policy topics for NOAK micro-reactors, including any related to safety, construction inspection, security, emergency preparedness, and environmental reviews.

The NRC staff is cognizant that some topics in this enclosure could be more broadly relevant to the deployment of all types of micro-reactors and other reactor technologies, such as small modular reactors and larger reactors. Although this enclosure does not explicitly address such situations, the NRC staff will account for them, including through further Commission engagement, as appropriate.

1.

Maximal Design Standardization in a Manufacturing License or Design Certification Overview Micro-reactor developers and potential applicants have indicated that they intend to pursue deployment of many micro-reactors of standard designs. In the context of this paper, a maximally standardized micro-reactor design has the following attributes and characteristics:

The design is a standard design as defined in the regulations in 10 CFR 52.1, Definitions, that state, [s]tandard design means a design which is sufficiently detailed and complete to support certification or approval in accordance with subpart B

[Standard Design Certification] or E [Standard Design Approvals] of this part, and which is usable for a multiple number of units or at a multiple number of sites without reopening or repeating the review.

The complete plant design is approved and provided finality in a design certification (DC) or manufacturing license (ML) or a combination of the two. (As discussed below, a standard design approval doesnt provide the same finality as a DC or ML and therefore would not achieve maximal design standardization.)

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The design uses bounding site parameters, including hazard parameters, that would make it suitable for licensing at deployment sites with a variety of site characteristics1 The design minimizes site-specific design features and interfaces between the reactor and the site (e.g., offsite power or water sources relied upon for safety functions).

The design of an individual reactor does not include departures from the approved design.2 The current regulatory framework provides several pathways for micro-reactor developers to seek approval of a standard design. The regulations in 10 CFR Part 52, Subpart B, Standard Design Certifications, sets forth the requirements and procedures applicable to Commission issuance of rules granting standard design certifications for nuclear power facilities. As stated in 10 CFR 52.47, Contents of application; technical information, the application for a standard design certification must, among other things, contain a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that construction conforms to the design and to reach a final conclusion on all safety questions associated with the design before the certification is granted.

The regulations in 10 CFR Part 52 Subpart E, Standard Design Approvals, sets out procedures for the filing, NRC staff review, and referral to the Advisory Committee on Reactor Safeguards of standard designs for a nuclear power reactor or major portions thereof. As stated in 10 CFR 52.137, Contents of applications; technical information, the application for a standard design approval must contain a final safety analysis report that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility, or major portion thereof. The regulations in 10 CFR 52.137 also state, [i]f the applicant seeks review of a major portion of a standard design, the application need only contain the information required by this section to the extent the requirements are applicable to the major portion of the standard design for which NRC staff approval is sought.

The regulations in 10 CFR Part 52, Subpart F, Manufacturing Licenses, sets out the requirements and procedures applicable to Commission issuance of a license authorizing manufacture of nuclear power reactors to be installed at sites not identified in the manufacturing license application. As stated in 10 CFR 52.157, Contents of applications; technical information in final safety analysis report, the application for a manufacturing license must, among other things, contain a final safety analysis report with a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that the manufacturing conforms to the design and to reach a final conclusion on all safety questions associated with the design.

The regulations in 10 CFR 52.63, Finality of standard design certifications,10 CFR 52.145, Finality of standard design approvals; information requests, and 10 CFR 52.171, Finality of manufacturing licenses; information requests, specify the finality provisions associated with 1

The regulations in 10 CFR 52.1 state, site characteristics are the actual physical, environmental and demographic features of a site. Site characteristics are specified in an early site permit or in a final safety analysis report for a combined license. The regulations in 10 CFR 52.1 also state, site parameters are the postulated physical, environmental and demographic features of an assumed site. Site parameters are specified in [an SDA], a standard design certification, or [an ML].

2 While an applicant may still propose departures from the approved design or postulated site parameters at a specific site, such departures would complicate the review of that particular application.

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To achieve maximal design standardization, a standardized design should minimize the likelihood that applicants for licenses that reference the standardized design will take departures or request design changes.4 In this respect, the use of bounding values for site parameters and minimizing site-specific design features and interfaces between the reactor and the site will contribute to achieving maximal design standardization. This can be accomplished in several ways. One approach would be to incorporate sufficient margins and flexibilities in the design characteristics, site parameters, terms and conditions, and approved design. Another approach would be to obtain NRC approval for several standardized models of a similar design where the design characteristics, terms and conditions, and design of each model would be optimized for a range of site parameters, such as seismicity, atmospheric conditions, or potential external hazards.

Other regulatory approaches can be used to support design standardization, although they do not provide a degree of finality that achieves maximal design standardization on their own. A first-of-a-kind CP/OL or COL can be referenced in a subsequent application using a design centered licensing review approach or it can be used to inform changes to a DC or ML to 3

In the 2007 final rule amending the regulations in 10 CFR Part 52 (72 FR 49392) the NRC stated, In light of the NRC's review and approval of a final design as part of issuance of a manufacturing license, the final rule provides a greater degree of finality to a manufacturing license as compared with a standard design certification.

Under §52.171(a)(1), the same degree of issue finality accorded to the certified design applies throughout the term of the manufacturing license.

4 The regulations in 10 CFR 52.63(b) and 10 CFR 52.171(b) include requirements related to departures from the design characteristics, site parameters, terms and conditions, or approved design in a DC or ML, respectively. In general, departures will complicate NOAK licensing proceedings and reduce efficiency by opening matters that were resolved in a DC or ML and therefore expanding the scope of the associated NRC staff review and hearing(s).

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NOAK Strategy Enhanced standardization is allowed under current regulations and would be consistent with the Commissions policy on standardization (Nuclear Power Plant Standardization (52 FR 34884))

which preceded the development of 10 CFR Part 52 and encouraged standardization to enhance plant safety, improve regulatory efficiency, and reduce the complexity and uncertainty of the regulatory process. As discussed in the main paper and this enclosure, the NRC staff intends to leverage maximal design standardization to streamline safety reviews, environmental reviews, operational program approvals, construction inspections, and other regulatory processes. This includes using maximal design standardization in a DC or ML to reduce the scope of the NRC staff safety review at the COL or CP/OL stage to verification of site characteristics and any other remaining site-specific matters. This is a critical aspect of achieving the greatest efficiency gains and shortening licensing timelines. As discussed in SECY-20-0093, the staff is also receptive to requests for exemptions from the existing regulations that would contribute to enhanced standardization and would evaluate such exemptions case by case using existing agency processes. The NRC staff remains committed to using risk insights and pursuing performance-based approaches, where justified, to improve regulatory processes, including licensing actions related to establishing standard designs and deploying micro-reactors of common standardized designs.

Pursuing maximal design standardization is voluntary and largely dependent on micro-reactor developers and applicants and their deployment models and licensing strategies. The NRC staff recognizes that stakeholders will seek to balance aspects of maximal design standardization, such as the level of conservatism in the design that supports bounding site parameters, with economic and operational considerations. In addition, stakeholders have expressed interest in having the flexibility to enhance designs over time as operational experience is gained with a particular design. The NRC staff can inform developers and individual applicants of the potential regulatory impacts (e.g., available licensing pathways, schedules, and estimated costs) of certain design decisions and changes to an approved standard design. Developers and applicants will have to weigh the potential benefits of maximal design standardization, such as enhanced NOAK licensing efficiency and schedule reduction and make their own decisions.

5 See Regulatory Issue Summary 2006-06, New Reactor Standardization Needed to Support the Design-Centered Licensing Review Approach, dated May 31, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML053540251).

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Next Steps Maximal design standardization is available to micro-reactor developers and potential applicants under the current regulatory framework and the NRC staff is not proposing any changes at this time. As stakeholders and the NRC staff gain experience with maximal design standardization and how it supports NOAK licensing, the NRC will assess whether additional guidance development or rulemaking could further enhance efficiency and engage the Commission, as appropriate.

The NRC staff will continue to engage with stakeholders on micro-reactor deployment models and licensing strategies that leverage maximal design standardization. The NRC staff encourages micro-reactor developers and potential applicants for a DC or ML to participate in preapplication engagement to foster a common understanding of the potential licensing pathways for approval of a standard design and to address design-specific regulatory matters early. Similarly, the NRC staff encourages preapplication engagement by potential applicants for a COL or CP/OL that intend to reference a standard design, especially in situations that may involve departures from the standard design.

2.

Grading the Level of Site Characterization Overview Several micro-reactor developers have thus far communicated that they plan to design micro-reactors using a set of postulated bounding site parameters6 so the reactors could be deployed at suitable sites throughout the U.S. without the need for customizing the designs for each site.

In accordance with 10 CFR 52.153, the applicants for the proposed deployment sites would need to obtain a COL (or CP) using the existing power reactor regulations and guidance for the identification and evaluation of the actual site characteristics necessary for determining the acceptability of the site and would verify that the site parameter values specified in the ML under 10 CFR 52.157(f)(19) (or in the DC) bound the corresponding site characteristic values. Site characterization information for CP or COL applications includes human-induced external hazards, meteorology, hydrology, geology, seismology, and geotechnical engineering considerations for a proposed site. The NRC staff expects that a graded approach can be used for both first-of-a-kind (FOAK) and NOAK CP or COL applications. Reductions in cost or review timeframe for the application of a graded approach are likely to depend on the individual site characteristics and design parameters.

The existing siting requirements in 10 CFR Part 100, Reactor Site Criteria, involve extensive site characterizations and investigations of the actual physical, environmental, and demographic features of a proposed site. Much of associated regulatory guidance7 were developed with large 6

Site parameters are the postulated physical, environmental, and demographic features of an assumed site and are specified in a DC or ML application. Site characteristics are the actual physical, environmental, and demographic features of a site that are specified in a CP or COL application. To avoid customizing or reanalyzing the design for each site, all the site parameter values specified in the DC or ML must bound the corresponding site characteristic values.

7 LWRs can use the guidance in Regulatory Guide (RG) 1.206, Applications for Nuclear Power Plants, (ML18131A181) and SRP Chapter 2.0, Site Characteristics and Site Parameters, (https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/ch2/index.html). Although RG 1.206 was developed for LWRs, the NRC

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LWR Edition, specifies that a COL application that does not reference an early site permit should include at least two consecutive years (and preferably 3 or more whole years, if available) of onsite meteorological data. This extensive site characterization is appropriate for large LWRs based on factors such as having large thermal power levels, large site footprints, large radionuclide inventories, and complex designs that use active safety systems.

However, based on the anticipated relative simplicity of micro-reactor designs, enhanced safety characteristics, and small site footprints, the NRC staff has determined that it is appropriate to develop a graded approach to site characterization for micro-reactors of a standard design.

NOAK Strategy In SRM-SECY-98-144, Staff Requirements - SECY-98-144 - White Paper on Risk-Informed and Performance-Based Regulation (ML24212A161)8, the Commission set its expectations for the NRC staff adopting risk-informed and performance-based approaches to support decision-making. In its Policy Statement on the Regulation of Advanced Reactors (73 FR 60612), the Commission recognized attributes that could assist in establishing the acceptability or licensability of a proposed advanced reactor design, including:

Simplified safety systems that, where possible, reduce required operator actions, equipment subjected to severe environmental conditions, and components needed for maintaining safe shutdown conditions. Such simplified systems should facilitate operator comprehension, reliable system function, and more straightforward engineering analysis.

Using concepts from SRM-SECY-98-144 and the policy statement, and the anticipated enhanced safety characteristics of micro-reactor designs, the NRC staff plans to develop guidance for a graded approach to site characterization for micro-reactors of a standard design.

The approach will align the level of the site characterization needed to be performed at the deployment sites with the demonstrated enhanced safety margins and lower radiological consequences of the approved standard micro-reactor design. This graded approach would staff also considers it to generally apply to non-LWRs. Non-LWRs implementing the Licensing Modernization Project methodology can use the guidance in RG 1.253, 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 (ML23269A222), DANU-ISG-2022-01, ARCAP Roadmap Interim Staff Guidance (ML23277A139), and DANU-ISG-2022-02, ARCAP Chapter 2 Site Information (ML23277A140).

8 Item 8, Risk-Informed, Performance-Based Approach, in SRM-SECY-98-144 states that a risk-informed, performance-based approach to regulatory decision-making combines the "risk-informed" and "performance-based" elements [previously discussed in the SRM], and applies these concepts to NRC rulemaking, licensing, inspection, assessment, enforcement, and other decision-making. Stated succinctly, a risk-informed, performance-based regulation is an approach in which risk insights, engineering analysis and judgment including the principle of defense-in-depth and the incorporation of safety margins, and performance history are used, to (1) focus attention on the most important activities, (2) establish objective criteria for evaluating performance, (3) develop measurable or calculable parameters for monitoring system and licensee performance, (4) provide flexibility to determine how to meet the established performance criteria in a way that will encourage and reward improved outcomes, and (5) focus on the results as the primary basis for regulatory decision-making.

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The applicants commitment to maximal standardization is essential for enabling the graded approach.

Figure 1: Conceptual Approach for Grading the Level of Site Investigation and Characterization Figure 1 presents a conceptual approach for grading the level of site investigation and characterization needed to be performed at the deployment sites for micro-reactors of a standard design. Starting at Boxes 1 and 2, the standard micro-reactor design with maximal standardization would be approved in an ML or DC including a set of postulated site parameters that bound the expected site characteristics. If the postulated site parameters include sufficient margin, the design could potentially be deployed at suitable sites throughout the U.S. without the need for customizing the design for each site. The design of the safety-related structures, systems, and components (SSCs) would account for the design basis external hazard levels (DBEHLs) associated with the set of postulated bounding site parameters. As part of an application for an ML or DC, the designer/applicant would also perform an assessment of each site characteristic to see if the design is insensitive to any of the site characteristics (meaning that the site characteristic cannot affect the design and operation of the standard micro-reactor design, (e.g., a buried reactor might be insensitive to high winds). If the design is insensitive to any of the site characteristics, then a corresponding postulated site parameter would not need to be established for the design. However, a lack of corresponding postulated site parameter would not obviate the need for an applicant to address the site characteristic as part of the CP or COL application or request an exemption.

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The COL (or CP) applicant that references the approved standard design would perform an assessment of the deployment site as shown in Box 3 to identify any site characteristics that are not applicable to the site (e.g. sites located far from a body of water would not be subject to tsunamis) and any new site characteristics that are unique to the site (e.g. close proximity to an active fault). The remaining steps in Boxes 4 to 6 would then be conducted for one site characteristic at a time. As shown in Box 4, the applicant would select the first site characteristic and perform an initial assessment using readily available information to determine the actual site characteristic value at the proposed deployment site. The applicant could use readily available information from previously accepted databases and models for regional and local site area characteristics, as appropriate. As shown in Box 5, the applicant would then determine the amount of margin (1) between the site parameter value approved in the standard micro-reactor design and the actual site characteristic value at the proposed deployment site, and (2) between the doses calculated for the postulated design basis accidents at the boundary of the exclusion area and the associated 25 rem dose reference value in 10 CFR 50.34. As shown in Box 6, the applicant would then use an NRC-approved graded approach matrix (presented conceptually in Figure 1, above, for the purpose of this discussion) to assess the margins and identify whether traditional, reduced, or minimal level of site characterization can be used for the selected site characteristic. The applicant would then complete the steps in Boxes 4 to 6 for every other site characteristic.

In developing guidance that would provide an NRC-approved graded approach matrix, the NRC staff plans to assess how to appropriately apply the graded approach for each site characteristic. The NRC staff will also develop appropriate values to define what is considered low, medium, and high for site parameter margins and exclusion area boundary (EAB) dose margin, as well as what is the appropriate level of site characterization needed for reduced and minimal levels of site characterization in this graded approach. The NRC staff is considering whether minimal site characterization could rely on the readily available information that was used during the initial site assessment step (Box 4 of Figure 1) using readily available information. In addition, depending on the specific site that was selected and the amount of margin between each site parameter used in the design and the corresponding site characteristic, the NRC staff is considering whether a minimal level of site characterization could potentially be used for all site characteristics if a low hazard site is selected. Conversely, if a site is selected with a high hazard related to a particular site characteristic, then there is the potential that the applicant will need to perform a traditional level of site characterization for that characteristic and a reduced or minimal level of site characterization for one or more of the other site characteristics.

As shown in Box 7, once all the site characteristic values have been determined, the applicant would submit the COL (or CP) application (incorporating by reference the ML, or the DC and ML, that approved the standard design) to the NRC staff for review. The COL (or CP) application would include a description and basis for (1) any site characteristics that are not applicable to the site, (2) any new site characteristics that are unique to the site, (3) the actual site characteristic values, (4) the level of margin between the site parameters and the site characteristics, and (5) the level of site characterization performed for each site characteristic.

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Regulations and Guidance The requirements related to siting of nuclear power reactors are included in NRC regulations in 10 CFR Part 100, 10 CFR Part 50, Domestic licensing of production and utilization facilities; and 10 CFR Part 52, Licenses, certifications, and approvals for nuclear power plants.

The regulations in 10 CFR 100.20, Factors to be considered when evaluating sites, identify the evaluation factors that the NRC staff will take into consideration in determining the acceptability of a site for a stationary power reactor. The regulations in 10 CFR 100.21, Non-seismic site criteria, identifies the criteria that applications for site approval for commercial power reactors shall demonstrate that the proposed site meets. The regulations in 10 CFR 100.23, Geologic and seismic siting criteria, identifies the principal geologic and seismic siting criteria that the NRC staff will take into consideration in its evaluation of the suitability of a proposed site and adequacy of the design bases established in consideration of the geologic and seismic characteristics of the proposed site.

The regulations in 10 CFR 100.20(c) and 10 CFR 100.21(d) require the evaluation of the physical characteristics of the site, including seismology, meteorology, geology, and hydrology, and the establishment of site characteristics such that potential threats from such physical characteristics will pose no undue risk to the type of facility proposed to be located at the site.

The regulations in 10 CFR 100.21(c) require the evaluation of site atmospheric dispersion characteristics such that normal effluent release limits and postulated accident dose reference values can be met. The regulations in 10 CFR 100.23(c) require that the geologic, seismic and engineering characteristics of the site and its environs be investigated in sufficient scope and detail to permit an adequate evaluation of the proposed site; provide sufficient information to support estimates of the Safe Shutdown Earthquake (SSE) ground motion; and permit adequate engineering solutions to actual or potential geologic and seismic effects at the proposed site. It further specifies that all geologic and seismic factors that may affect design and operation of the proposed nuclear power plant must be investigated, irrespective of whether such factors are explicitly included (e.g., volcanic activity). The regulations in 10 CFR 100.23(d) require that the geologic and seismic siting factors that must be considered for design include a determination of the SSE ground motion for the site (including the uncertainty in the development of the SSE ground motions), the potential for surface tectonic and nontectonic deformations, the design bases for seismically induced floods and water waves, and other design conditions, such as soil and rock stability, liquefaction potential, natural and artificial slope stability, and remote safety-related structure siting.

The regulations in 10 CFR Part 50 and 10 CFR Part 52 address the design of nuclear reactors and take into consideration the site criteria contained in 10 CFR Part 100 in determining whether the proposed facility can be constructed and operated at the proposed site location without undue risk to the health and safety of the public. Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 contains the general design criteria (GDC) that establish the minimum requirements for the Principal Design Criteria (PDC) for LWRs.9 General Design Criteria 2, Design bases for protection against natural phenomena, requires that SSCs 9 The GDC are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance for establishing the PDC for such other units. RG 1.232, Developing Principal Design Criteria for Non-Light Water Reactors, provides guidance for establishing PDC for non-LWRs (ML17325A611) and includes proposed design criteria for non-LWR technologies that are the same as GDC 2.

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Chapter 2.0, Site Characteristics and Site Parameters, of the SRP provides guidance to staff to ensure that the appropriate actual site characteristics are identified at the proposed site for early site permit and COL applications submitted under 10 CFR Part 52. Section 2.1.1, Site Location and Description (ML070550023), to Section 2.1.3, Population Distribution (ML070550028), and Section 2.2.1, Identification of Potential Hazards in Site Vicinity (ML070460330), to Section 2.2.3, Evaluation of Potential Accidents (ML070460336), of the SRP provide guidance to the staff on population density and human-induced external hazards, including hazards from nearby facilities in the site vicinity. Sections 2.3.1, Regional Climatology (ML063600393), to Section 2.3.5, Long-Term Atmospheric Dispersion Estimates for Routine Releases (ML070730713), of the SRP provide guidance to the NRC staff in the field of meteorology, considering regional climatology and local meteorology, onsite meteorological monitoring and atmospheric dispersion estimates. Section 2.4.1, Hydrological Description (ML070100646), to Section 2.4.12, Groundwater (ML070730443), of the SRP provide guidance to the staff in the field of hydrology, considering several flood-causing mechanisms such as riverine based flood and upstream dam failure, local intense precipitation, tsunami, storm surge, seiche, ice dams and jams, channel migration, and groundwater. Sections 2.5.1, Basic Geologic and Seismic Information (ML13316C067), to Section 2.5.5, Stability of Slopes (ML13316C068), of the SRP provide guidance to the staff in the field of geology and seismology, considering regional and local geologic and seismological hazards as well as onsite geologic and seismic hazards.

There are a number of regulatory guides (RGs) that are applicable to site characterization information. In addition to the RGs listed below, the NRC maintains numerous hazard-specific RGs that provide guidance on how to meet the applicable regulatory requirements and are a useful starting point in considering how to apply the graded approach to CP or COL applications for a standard micro-reactor design. Some of these RGs are listed below.

RG 1.206, Rev. 1, Application for Nuclear Power Plants (ML18131A181), applies to LWR technology, but is also generally applicable to other types of power reactors (e.g.,

non-LWRs).

RG 4.7, General Site Suitability Criteria for Nuclear Power Stations (ML23348A082) provides guidance to staff on meeting the requirements in 10 CFR Part 100 for the major site characteristics considered in determining the suitability of sites for commercial nuclear power stations.

RG 1.253, 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 (ML23269A222), DANU ISG-2022-01, ARCAP Roadmap Interim Staff Guidance (ML23277A139), and DANU-ISG-2022-02, ARCAP Chapter 2 Site Information (ML22048B541), provide site characterization guidance to non-LWR applicants implementing the Licensing Modernization Project methodology described in Nuclear Energy Institute (NEI) report NEI 18-04, Risk-Informed Performance-Based Guidance for NonLight Water Reactor Licensing Basis Development (ML19241A336), as endorsed in RG 1.233, Guidance for a Technology-

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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).

RG 1.208, A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion (ML070310619), endorses ASCE 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, for determining the site-specific SSE or ground motion response spectra (GMRS).10 The NRC staff have already considered how to appropriately scale site characterization information for advanced reactors with a smaller radiological risk profile. An important example is the graded siting approach for seismic hazard characterization referred to as the SSHAC

((Senior Seismic Hazard Analysis Characterization) approach, which is endorsed in RG 1.208.

Additional examples are the incorporation of engineering judgement in the assessment of volcanic (RG 4.26, Volcanic Hazards Assessment for Proposed Nuclear Power Reactor Sites (ML23167A078)) and flooding (DG-1290, Revision 3 to RG 1.59, Design-Basis Floods for Nuclear Power Plants (ML19289E561)) hazards, which, if final, would allow an applicant to determine a maximum magnitude hazard and then consider the effect of that hazard on the specific reactor design. The graded approach for standard micro-reactor designs will similarly consider alternative approaches to site characterization, such as those used for lower risk profile facilities like research reactors. The NRC staff will also consider how to apply a graded approach for standard micro-reactor designs to be installed and operated at a site subject to a previous site characterization study, such as the site of an existing micro-reactor or other NRC-licensed facility.

Exemptions Based on an initial examination of the current regulations, the NRC staff does not anticipate that exemptions will be needed to implement a graded approach to site characterization for micro-reactors of a standard design. In general, 10 CFR 100.20 and 10 CFR 100.21 state requirements that are to a large extent performance based, which supports implementation of the graded approach. However, because 10 CFR 100.20(c)(3) and 10 CFR 100.23 include some more prescriptive requirements, the NRC staff will further assess the need for exemptions as it develops the guidance for the graded approach to site characterization for micro-reactors of a standard design. For example, 10 CFR 100.20(c)(3) includes a prescriptive requirement that factors important to hydrological radionuclide transport must be obtained from on-site measurements. In addition, 10 CFR 100.23(c) requires that investigations of sufficient scope and detail to determine the subsurface soil and rock properties and the geological characteristics of local and regional faults to determine the SSE ground motion for the site. The NRC staff will also assess whether implementation of the related guidance for the current operating fleet provides any useful examples for how NRC interprets the siting regulations.

10 The NRC staff is updating RG 1.208 to endorse ANS 2.27-2020, Criteria for Investigations of Nuclear Facility Sites for Seismic Hazard Assessments, and ANS 2.29-2020, Probabilistic Seismic Hazard Analysis, to provide for a graded approach for site investigation and seismic hazard characterization. This activity is not yet complete and prospective applicants that wish to use these standards prior to the issuance of the Revision to RG 1.208 should discuss their plans with the NRC staff during the preapplication phase.

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Next Steps The NRC staff plans to develop guidance for a graded approach to site characterization for micro-reactors of a standard design that aligns the level of the site characterization needed to be performed at the deployment sites with the demonstrated enhanced safety margins and lower radiological consequences of the standard micro-reactor design. A review team of NRC staff experts representing human-induced external hazards, meteorology, hydrology, geology, seismology and geotechnical engineering are beginning to look for ways to adapt existing regulations and guidance for site characterization of LWRs for use in a graded approach for standard micro-reactor designs fabricated under a ML and deployed to an approved site. The NRC staff plans to first release the draft guidance as a white paper and conduct public meetings to facilitate the receipt of early stakeholder feedback.

3.

Deployment Site Emergency Preparedness Overview Per 10 CFR 50.47, Emergency Plans, the NRC will only issue an OL, COL, or early site permit (ESP) that includes complete and integrated emergency plans under 10 CFR 52.17(b)(2)(ii),

after a finding is made that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Whether an applicant chooses to use 10 CFR 50.47(b) and Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, to 10 CFR Part 50 or the requirements in 10 CFR 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities, the applicant must conduct an initial exercise to demonstrate compliance with the emergency plan within 2 years before the issuance of an operating license for the facility described in the license application for 10 CFR Part 50, or within two years before the scheduled date of initial fuel load for a combined license issued under 10 CFR Part 52. For 10 CFR Part 52 COL applicants, 10 CFR 52.97(b) requires that the site-specific COL requirements will identify the inspections, tests, and analyses, including those applicable to emergency planning, that the licensee shall perform, and the acceptance criteria that, if met, are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations.

Guidance The NRC employs a graded approach to emergency preparedness (EP) in which requirements are set commensurate to the radiological risks and hazards of the facility. Micro-reactor applicants can use this risk-informed approach to EP to establish emergency plans and emergency response capabilities. RG 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities, dated November 2023 (ML23226A036) provides guidance for establishing a commensurate level of EP using a risk-informed, performance-based and technology-inclusive approach for applicants complying with 10 CFR 50.160. The NRC staff is also developing an interim staff guidance (ISG) for EP reviews that would facilitate consistency and predictability of technical staff reviews for applicants complying with 10 CFR 50.160.

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RG 1.242 was written to be technology inclusive, with the expectation that specific methods for complying with EP regulations would be developed for different designs. Based on current understanding of design characteristics of micro-reactors and available source term information, staff is considering development of additional guidance to facilitate efficient licensing approaches to EP, including sizing of the emergency planning zones (EPZ) specifically for micro-reactors. Such guidance could be comparable to the approach to sizing EPZs for research and test reactors, as described in ANSI/ANS-15.16-2008, Emergency Planning for Research Reactors. For example, additional guidance on standard approaches to EP for micro-reactors (less than several tens of megawatts) could be developed and incorporated into a future revision of RG 1.242. Consistent with the discussion in Enclosure 1, the staff will also continue to assess additional ways to implement efficiencies in its reviews for specific micro-reactor technologies and standard EP approvals at the DC/ML stages.

Emergency Plan Implementing Procedures Some elements of emergency planning have associated licensing milestones that would need to be considered when developing licensing strategies for NOAK micro-reactors. For example, paragraph V of Appendix E to 10 CFR Part 50 requires detailed emergency plan implementing procedures (EPIPs) be submitted no less than 180 days before the scheduled issuance of an operating license, or the scheduled date for initial fuel loading for a COL under 10 CFR Part 52.

While the EPIPs are submitted for NRC review but do not require approval, the required timing for submittal of detailed procedures could challenge licensing schedules for some micro-reactor applicants. Standardized detailed EPIPs could be reviewed to the extent practicable and could include placeholders for site-specific information for NOAK applicants.

Use of 10 CFR 50.47(b) and Exemptions Applicants could further choose an approach to comply with the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50 instead of 10 CFR 50.160 and seek exemptions from certain requirements based on differences in design characteristics and low radiological risks. If potential applicants indicate an intent to take this approach, staff would evaluate the need for guidance development to facilitate more efficient exemption request reviews. This could be analogous to similar guidance issued in 2015, NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants (ML14106A057), which was written to support more efficient and standardized staff reviews of exemption requests for decommissioning plants.

Next Steps The staff will continue to develop an ISG for the review of applications under 10 CFR 50.160 and continue to assess additional ways to implement efficiencies within RG 1.242. As appropriate, the NRC staff will develop and issue additional risk-informed guidance and strategies that are specific to the application of 10 CFR 50.160, and that may be needed to enhance the efficiency of licensing and oversight for microreactors and multiple-reactor deployment models.

The staff is also considering conducting limited research to further identify potential bounding approaches for addressing standardized EP approvals of lower-risk reactors (e.g., not greater than several tens of megawatts) to maximize staff licensing and oversight efficiency. During

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4.

Deployment Site Security Overview All commercial nuclear power plants (NPPs) in the United States are currently licensed under 10 CFR Part 50 or 10 CFR Part 52 The security programs that apply to NPP applicants and licensees are specified in 10 CFR 73.55, Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. A high-level overview of several security programs can be found in Enclosure 1.

An applicant or licensee of an NPP may seek an exemption or an alternative measure to one or more of the physical security requirements in 10 CFR 73.55. The process to do so is described in 10 CFR 73.5, Specific Exemptions and 10 CFR 73.55(r), Alternative Measures. In either case, an applicant or licensee is required to obtain NRC approval before departing from the regulations. Barring exigent circumstances, either process can take a year or longer to complete.

As the nuclear industry is exploring new reactor designs, the NRC is developing draft proposed rulemakings to consider the technological advancements associated with those designs. Two current draft proposed rulemakings are Alternative Physical Security Requirements for Advanced Reactors, (SECY-22-0072, ML21334A003) which would allow small modular reactors and advanced reactor applicants to implement physical security alternative approaches, and the Nuclear Energy Innovation and Modernization Act directed, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors, (SECY 0021, ML21162A093) which would create an optional framework for all new reactor applicants in a new 10 CFR Part 53. Under a proposal for consideration in 10 CFR Part 53, a licensee could be exempt from defending against the design basis threat of radiological sabotage if it demonstrated that the facility met a low dose consequence threshold. Proposed 10 CFR 73.120 in 10 CFR Part 53 would provide flexibility by making available an alternate approach, commensurate with risk and consequence to public health and safety, for 10 CFR Part 53 applicants that can demonstrate in an analysis that the offsite consequences of a postulated event will meet the criterion defined in 10 CFR 53.860(a)(2)(i). The analysis must assume that licensee mitigation and recovery actions, including any operator action, are unavailable or ineffective. Under this proposed approach, should an applicant for a commercial nuclear reactor license demonstrate, pursuant to 10 CFR 53.860(a)(2)(i), that an offsite release would not exceed doses defined in the safety criteria of 10 CFR 53.210, the applicant may implement the access authorization program requirements under 10 CFR 73.120, instead of the requirements under 10 CFR 73.55 or 10 CFR 73.100, 10 CFR 73.56, and 10 CFR 73.57.

Other requirements to consider for streamlining a licensing review are the siting criteria in 10 CFR 50.34(a)(1)(i) and 10 CFR 100.21(f) which state, Each application for a construction permit (CP) shall include a Preliminary Safety Analysis Report. The minimum information to be included shall consist of. A description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. Special attention

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Additionally, as required by 10 CFR 52.97(b), the Inspections, Tests, Analysis, and Acceptance Criteria (ITAAC) identified in the Combined Operating License (COL) are necessary and sufficient, when successfully completed by the licensee, to provide reasonable assurance that the facility has been constructed and will operate in conformity with the COL, the provisions of the Atomic Energy Act of 1954, as amended (AEA), and the Commission's rules and regulations. ITAAC as described above are not requirements for an OL under 10 CFR Part 50. It will be necessary for the NRC staff to verify completion of ITAAC in regard to the security design to make a finding for authorization to operate under 10 CFR 52.103(g) or to verify substantial completion of construction for issuance of an OL under 10 CFR 50.56 and 10CFR 50.57(a)(1).

The FOAK micro-reactor license application bears the most risk, and therefore would require more agency resources for processing the application. This could result in longer timelines associated with technical, regulatory, and legal reviews and increased management oversight, to reach a regulatory decision for the application. For never-before-reviewed designs, preapplication engagement with the NRC staff will help to identify technical hurdles and information needed in the application. Developers spend substantial resources preparing a FOAK micro-reactor application and must also cover the NRCs staff and user fees for these activities.

NOAK Strategy:

In addition to the information discussed in Enclosure 1, once the NRC reviews and the Commission approves security plans for a FOAK micro-reactor design, the NOAK micro-reactor applications may be streamlined. During the review of the NOAK micro-reactor security plans, any changes identified would need to be re-evaluated, and the application process would potentially have to start from the beginning through an additional review of information that contradicts the previously approved FOAK security plans. The physical security for an approved FOAK micro-reactor is designed to mitigate the risk of a potential malevolent act. If the manufacturer makes a change, for example a change in the fuel loading, this could increase the risk profile of the facility requiring a subsequent change in physical security. All significant security parameters including site-specific considerations that are relevant to standardization being sought will be reviewed and approved by the NRC in reviewing a COL application for an FOAK reactor. The FOAK micro-reactor COL review can then be used to inform NRC reviews for follow on NOAK micro-reactor applications.

Operational programs utilized by a licensee, such as security, play a critical role in the overall efficiency, cost, and timeline of NOAK micro-reactors. Enhanced efficiency can be achieved through the standardization of security protocols.

Staff also note that the more that security by design can be incorporated upfront into the reactor design process, potential security concerns can be addressed during the design and incorporated into the additional designs that would follow for the NOAK micro-reactor design.

This early engagement can lower the ongoing costs of security for the facility.

However, there are certain challenges to this strategy. The requirement of 10 CFR 100.21(f) requires that the site characteristics must permit adequate security plans and measures to be

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Special Nuclear Material The NRC staff anticipates that most of the advanced reactors will be utilizing special nuclear material (SNM) of moderate strategic significance (Category (CAT) II) as defined in 10 CFR 73.2, Definitions. The physical security regulations for CAT II material have not been updated since 1979 and do not address the current threat environment. As a result, CAT II physical security has been addressed on a case-by-case basis. Consistent with this approach, the Commission, in SRM-SECY-18-0063, approved supplemental security measures for physical protection at fixed sites and in transit for Category II quantities of SNM to produce molybdenum (Mo)-99.

Next Steps The NRC staff is continuing to update the regulatory framework. For instance, the NRC staff is considering approaches for addressing SNM concerns to help increase regulatory certainty associated with the physical security of CAT II SNM facilities.

5.

Streamlined Licensing Process Overview Micro-reactor stakeholders have indicated that widespread deployment of micro-reactors could result in hundreds of license applications being submitted to the NRC for review and approval over the next decade and beyond. Submission and processing of applications for commercial nuclear power reactors are information-intensive activities that involve many administrative steps in addition to safety and environmental reviews by the NRC staff.11 In the past, most applications for CP/OLs or COLs have been for custom designs and the NRC staff practice has been to create the licensing documents for each application submitted under 10 CFR Part 50 or 10 CFR Part 52 as the processing and review of the application proceeds.

Each application for a CP/OL under 10 CFR Part 50 will have to include the information required by the regulations in 10 CFR 50.33, Contents of applications; general information, and 10 CFR 50.34, Contents of applications; technical information. Each application for a COL under 10 CFR Part 52 will have to include the information required by the regulations in 10 CFR 52.77, Contents of applications; general information, 10 CFR 52.79, Contents of applications; technical information in final safety analysis report, and 10 CFR 52.80, Contents of applications; additional technical information. In addition, applications for licenses under both 10 CFR Part 50 and 10 CFR Part 52 must include the information specified in 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, to support the NRC staff reviews required by the National Environmental Policy Act (NEPA). Much of the information required by 10 CFR Part 50 and 10 CFR Part 52 is the same 11 This information topic focuses on the safety aspects of licensing. The vote topic on alternative environmental reviews in the main paper and supporting information in Enclosure 2, Environmental Reviews for Nth-of-a-Kind Micro-Reactors, focuses on the environmental aspects of license applications and related NRC staff review.

Although this paper treats the safety and environmental aspects of licensing separately, they would likely be performed in parallel in a coordinated manner under the proposed licensing approach.

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Application, including the FSAR Supplemental information Notification of scheduled date for initial fuel loading required by 10 CFR 103(a)

Notifications related to ITAAC required by 10 CFR 52.99(a) and (c)

Hearing-related information The NRC staff will have to process each application in accordance with the administrative procedures required by the AEA and the relevant regulations in 10 CFR and perform the verifications and reviews necessary to make required determinations about adequate protection of public health and safety and the common defense and security. The current regulatory framework requires the NRC staff to produce many documents throughout the CP/OL and COL licensing processes and these typically include the following:

Licensing documents NRC staff safety evaluation NRC staff environmental review and record of decision Letter report of the ACRS Construction permit Operating license Combined license Technical specifications Notices published in the Federal Register Notice of Acceptability for docketing and availability of the application Notice of mandatory hearing Notice of opportunity to request a hearing Notices related to NRC NEPA documents Notice of issuance of a construction permit, operating license, or combined license Notice of the NRC staff's determination of the successful completion of inspections, tests, and analyses (for a combined license)

Notice of intended operation required by 10 CFR 52.103(a) (for a combined license)

Notice of a finding that the acceptance criteria in a COL are met Notices required by AEA sec.182c.12 12 Section 182c. of the AEA states, The Commission shall not issue any license under section 103 for a utilization or production facility for the generation of commercial power until it has given notice in writing to such regulatory

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NOAK Strategy The NRC staffs strategy for efficient processing of many CP/OL or COL applications referencing a standard micro-reactor design approved in a DC or ML includes three main aspects: design-specific templates for the contents of applications, an online system for application submission and processing, and automated licensing document templates. The objective of the approach is to leverage technology and standardization to the greatest extent practicable to automate many of the processes that are currently done manually by the NRC staff and to focus the NRC staff review of an application on site-specific matters and new information not approved in the DC or ML (or both) referenced in the application.13 This approach will have the greatest benefit for applications that reference a common maximally standardized design (i.e., a design approved in a DC or ML and without application-specific departures) and previously approved standardized operational programs. The cumulative benefits will scale with the number of applications and the NRC staff anticipates that the resources needed to process a NOAK license application (not including the upfront cost of establishing the application template) could be significantly reduced compared to those that have been typical for processing one-of-a-kind applications in the past.

Under this approach, an applicant for a DC or ML or a DC sponsor or ML holder would request NRC adoption of a design-specific template for the standard contents of an application for a COL or CP/OL for a reactor that references the design approved in the DC or ML. This could potentially be done in parallel with the DC or ML proceeding, through a topical report, or using another appropriate vehicle. Alternately, a person who intends to apply for many licenses referencing a common standard design could develop the template. The NRC would review the template and the proposed standard technical contents to ensure that it encompasses all of the information that the NRC staff would need to make the findings related to the adequacy of the design necessary to issue a COL or CP/OL referencing the standard design. The template would include placeholders for site-and applicant-specific information, including the reactor location and site characteristics, external hazards, other site-specific information (e.g.,

information supporting emergency preparedness), design information not resolved in the DC or ML, information related to on-site construction, operational programs, information related to decommissioning planning, and other information.

The second aspect of this approach is an online system that an applicant would use to submit the application and subsequent information, and the NRC staff would use to process the application. The system would allow applicants to register and provide much of the applicant-specific information required by 10 CFR Part 50, 10 CFR Part 52, and related regulations, such as information required by 10 CFR 50.33, Contents of applications; general information. The design-specific template for the standard contents of an application would be used to generate an electronic application form. This form would include fields for providing the necessary agency as may have jurisdiction over the rates and services incident to the proposed activity; until it has published notice of the application in such trade or news publications as the Commission deems appropriate to give reasonable notice to municipalities, private utilities, public bodies, and cooperatives which might have a potential interest in such utilization or production facility; and until it has published notice of such application once each week for four consecutive weeks in the Federal Register, and until four weeks after the last notice.

13 The NRC staff is not proposing to use artificial intelligence to perform safety reviews of applications. Rather, the NRC staff is proposing to automate the process of taking data from one document (information in an application) and placing it into other documents that would otherwise be done manually.

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The form would also request supporting information, such as site characterization information, necessary for the NRC staff to perform independent review and verification. As processing of the application proceeds, the system would also prompt applicants and licensees to provide additional information, such as information related to ITAAC notifications required by 10 CFR 52.99. The system would include features to satisfy existing requirements for filing applications, such as electronic signatures, oath or affirmation, and requests to withhold proprietary information from public disclosure.

The online system would facilitate efficient NRC staff processing of the application by automating many administrative functions, serving as a centralized hub for the safety review, and streamlining the review and issuance of licensing documents. Upon submission of an application, the system could automatically check that the application is complete (i.e., all application fields contain valid values and supporting information is provided), assign a docket number, and create consolidated public and non-public versions of the application. The system would generate but not immediately issue the notice of acceptability for docketing and the notice of mandatory hearing using the automated templates described below. The NRC staff would perform quality checks and route the documents for necessary management and legal reviews, after which the system would be used to distribute the application and send the notices to the Federal Register for publication.14 The system would populate the template for the NRC staff safety evaluation with information specific to the applicant and the referenced common standard design and notify assigned reviewers to perform confirmatory checks and verification of the adequacy of supporting information. Similarly, the system would generate and populate other licensing documents as the review proceeds such as the license and required Federal Register notices using information from the applicants system registration, application, supplements provided by the applicant, NRC staff safety evaluation, and hearing documentation. A major benefit of the system would be that all information related to processing the application would be centrally stored in a way that it would be available for automatic propagation throughout licensing documents as the review of an application proceeds.

The third aspect of this approach is the use of automated templates for NRC licensing documents. In this context, automation refers to the ability of the online system to automatically populate the templates with information stored in the system. For the most part, the NRC staff anticipates that these templates could be prepared generically to apply to all applications for micro-reactors that reference common maximally standardized designs. Some templates, such as the template for the NRC staff safety evaluation, would likely be further tailored to each specific design. Also, the suite of templates would need to be tailored to the means of approval of the standard design (DC or ML) as well as the licensing pathway for deployment of the reactor (CP/OL or COL). In developing the templates, the NRC staff would strive to maximize the use of standardized language in a way that supports inclusion of design-and applicant-specific information without the need for lengthy custom narrative descriptions. The goal would be to reduce the resources needed for the NRC staff to adequately document its safety review and for the required management and legal reviews. This approach would necessarily involve great attention to detail in the creation of the FOAK licensing documents that would form the 14 Although not discussed in detail here, the NRC staff would assess and adjust as appropriate its internal processes for management and legal review considering the anticipated repetitive nature of processing nearly identical applications for micro-reactors of a common design and the risk profile of micro-reactors.

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The streamlined processing of applications and licensing documents described in this information topic would provide a potential approach to address the direction in Section 208(b)(2), (3), and (4) of the ADVANCE Act. Specifically, the use of automated standardized templates for licensing documents would reduce redundancies and enhance efficiency (Sec. 208(b)(2)) compared to the NRC staffs current practice of developing the suite of licensing documents for each application and manually populating the information. The online system described in this information topic would also help to consolidate review phases, reduce transitions between review teams, and establish integrated review teams (Sec. 208(b)(3) and (4)) by serving as a centralized information repository and application processing environment.

It is possible that the online system could also be used to facilitate information exchange related to hearings and coordination with other agencies involved in NRCs licensing processes, such as the Department of Homeland Security and the Federal Emergency Management Agency, consistent with the concepts for intra-agency cooperation on environmental reviews specified in Section 506 of the ADVANCE Act.

Next Steps The NRC staff will continue to engage with stakeholders, including micro-reactor developers and potential applicants, to better understand their licensing strategies and deployment models and how increased use of electronic processing and automation can enhance regulatory efficiency. Based on the outcomes of these engagements, the NRC staff will identify potential software solutions for establishing an online system for submission and processing of applications, considering existing NRC licensing support systems, experience in other industries, and the NRCs experience with online applications for other regulatory activities, such as relief requests and materials licenses. The NRC staff will also consider the need to develop guidance for applicants and licensees to propose and the NRC staff to review design-specific templates for standard contents of applications. Although the NRC staff anticipates that an online system for applications and the related NRC staff safety review can be implemented consistent with the current regulations and Commission policies, the NRC staff will assess the need for further Commission engagement as development proceeds.

6.

Construction Inspection Overview of the topic While the Construction Reactor Oversight Process (cROP) provided effective oversight for construction of large light water reactors (LWRs) at the Vogtle site, given the expected diversity of advanced reactor projects and increase in offsite fabrication and assembly of safety-significant SSCs, it is the staffs view that a fresh look at construction oversight is warranted in the development of the Advanced Reactor Construction Oversight Program (ARCOP). To communicate to the Commission the plan to develop the ARCOP, the staff issued SECY-23-0048, Vision for the Nuclear Regulatory Commissions Advanced Reactors Construction Oversight Program (ML23061A086), dated June 6, 2023. The NRC staffs ARCOP vision looks to build on its construction oversight experience while remaining

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risk-informed: uses facility risk insights to define the scope of inspection; performance-based: adjusts oversight based on performance of licensees and suppliers; technology-inclusive: covers the full spectrum of advanced reactor technologies being considered for NRC licensing; scalable: uses a graded approach to inspection efforts commensurate with a facilitys public health and safety risk; informed: applies construction oversight experience and leverages lessons from past and current NRC inspection programs and other external sources; comprehensive: provides for oversight of all activities that are significant to construction quality (procurement, manufacturing, and onsite construction), security programs, and operational readiness; and, innovative: leverages new inspection tools and approaches, such as hybrid inspections, to enhance efficiency and effectiveness.

Since this vision was communicated to the Commission, the staff has continued ARCOP development and is in the process of developing guidance documents to ensure full implementation readiness to oversee advanced reactor construction and manufacturing activities in support of a finding on whether or not acceptance criteria are met under 10 CFR 52.103(g) or to verify substantial completion of construction for issuance of an operating license under 10 CFR 50.56, Conversion of construction permit to license; or amendment of license, and 10 CFR 50.57(a)(1). To support development of the program, the NRC staff has conducted extensive outreach with stakeholders, including hosting a series of public workshops to explore, in part, ideas on how best to scope the inspection program in a risk-informed, performance-based, technology inclusive, and as importantly, scalable manner. The staff will continue to develop these approaches and will communicate them to the Commission when they are more fully developed.

NOAK Strategy While the ARCOP can serve the oversight role needed for manufacturing and site deployment of first-of-a-kind technologies, the decision-making processes that are being developed will also enable adjusting oversight for NOAK circumstances. Specifically, application of the ARCOP inspection scoping methodologies will identify a risk-informed, performance-based inspection footprint for FOAK construction and deployment that will include the inspection necessary to verify that licensee/manufacturer performance within the ARCOP framework provides reasonable assurance that reactors of a particular design are being built and will operate in accordance with the designs licensing basis. As experience is gained in manufacturing and deployment of a given technology into future deployments, it is expected that the construction inspection footprint would be adjusted to reflect this gained experience. In this way, the program is scalable to reduce the amount of inspection needed to the minimum required to meet program objectives. The amount of scalability that can be achieved would be largely affected by

Preliminary White Paper - Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, Enclosure 3 (September 2024) 22 Preliminary White Paper - Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, Enclosure 3 (September 2024) the amount of standardization between FOAK and subsequent deployments. Significant changes to the technology itself or construction/manufacturing techniques could limit the ability to inform the scope of subsequent deployments by incorporating past experience, which is key to reducing the timeframe needed to perform inspections to enable rapid deployment of these technologies.

Next Steps The staff expects to communicate the details of the ARCOP in a separate paper to the Commission in December 2024