Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors

Subpart F provides the requirements for the operations phase of a commercial nuclear plant to ensure that the safety criteria in subpart B are satisfied throughout the plant's lifetime and during all modes of normal operation and unplanned events. Section 53.700 provides the general organization and overall objectives of subpart F, which are to establish requirements during operations for (1) plant SSCs; (2) personnel; and (3) plant programs.

Section 53.710 provides the requirements for maintaining capabilities, availability, and reliability of SSCs to demonstrate compliance with the safety criteria and design requirements for unplanned events that are described in subparts B and C. The basic structure of this section is that measures for SR SSCs are provided by TS and measures for NSRSS SSCs are required to be addressed with licensee-controlled documents and procedures.

The general content and control of TS under part 53 are similar to the requirements in part 50. The requirements for TS include limits on the inventories of radioactive materials, plant operating limits, and specific requirements for each SR SSC, including limiting conditions for operation (LCO) and required surveillances. The requirements for TS also include a section on important design elements, which is similar to design features in Sec. 50.36, and a section for administrative controls. A provision addressing the development and submittal of TS to address decommissioning activities is also included in subpart G.

The requirements for TS under part 53 do not carry over safety limits or associated limiting safety system settings from Sec. 50.36, which contains TS requirements for operating reactors under parts 50 and 52. As discussed in SECY-18-0096, systematic assessments and more mechanistic approaches to evaluating source terms support an alternative approach to establishing barrier-based safety limits. An example provided in that paper is a comparison of: (1) the traditional specified acceptable fuel design limits (SAFDL) that support protecting a specific barrier from potential failure mechanisms (e.g., departure from nucleate boiling to protect fuel cladding); and (2) the specified acceptable system radionuclide release design limit (SARRDL) concept, which limits the possible increase in circulating radionuclide inventory during normal operations or an AOO as part of an integrated or “functional containment” approach. Additional discussion of the use of SARRDL in the design and licensing of advanced reactors is provided in RG 1.232. The SARRDL could be addressed as an operating limit within this construct of requirements for TS. In cases, such as LWRs, where a SAFDL approach might be used as part of a mechanistic approach to meeting the design and analysis requirements in subpart C, the associated functional design criteria in Sec. 53.410 and TS under Sec. 53.710(a) define similar requirements as those provided by the safety limit and limiting safety system setting requirements in Sec. 50.36.

The requirements for TS under part 53 do not include specific criteria for identifying when LCOs must be established (i.e., do not include an equivalent to Sec. 50.36(c)(2)(ii)). Instead, consistent with subparts B and C, the TS requirements in subpart F of part 53 define TS LCOs as providing limits on SR SSCs. The SR SSCs protect against DBAs to demonstrate compliance with the safety criteria in Sec. 53.210. In the construct for part 53, risk-significant SSCs are addressed through a combination of TS for SR SSCs and establishment and monitoring of performance standards for NSRSS SSCs.

In addition to addressing TS for SR SSCs, Sec. 53.710 requires appropriate control measures be developed and implemented for NSRSS SSCs. Examples include appropriate surveillances and controls established through reliability assurance programs. Configuration management and other special treatments provide that the capabilities, availabilities, and reliabilities of NSRSS SSCs are maintained consistent with the underlying risk assessments while providing flexibility to licensees through maintaining the management functions within licensee-controlled programs. Controls on NSRSS SSCs are appropriate as part of the overall performance-based approach within part 53. Special treatments beyond those defined for their SR functions may also be warranted for SR SSCs to reflect their role in meeting the safety criteria in Sec. 53.220 and the evaluation criteria in Sec. 53.450(e). The performance objectives for NSRSS SSCs reflect that the comprehensive risk metrics and related risk performance objectives established under Sec. 53.220 may involve assessing and averaging the risks over a defined period (e.g., plant year) and do not constitute a real-time requirement that must be continuously demonstrated by the licensee. The controls under Sec. 53.710(b) justify changes in part 53 from the traditional or deterministic approaches in parts 50 and 52 in areas such as replacing the single-failure criterion with a probabilistic reliability criterion (see SRM-SECY-03-0047, “Policy Issues Related to Licensing Non-Light-Water Reactor Designs,” dated June 26, 2003). This approach can also support the incorporation of risk insights and analytical margins to gain operational flexibilities in areas such as siting and staffing requirements described in subsequent sections of subpart F.

Section 53.715 provides the requirements for developing and implementing a program to do the following: (1) control maintenance activities; (2) take appropriate corrective action when performance issues are identified; (3) conduct routine evaluations of effectiveness; and (4) assess and manage risks resulting from maintenance activities. These requirements are similar to those included in Sec. 50.65 (maintenance rule), including the need to assess and manage the increase in risk that may result from the maintenance activities. While, for the maintenance rule, specific criteria must be developed to capture both SR and non-SR but otherwise important SSCs, Sec. 53.715 covers SR SSCs and NSRSS consistent with other subparts in part 53.

Section 53.720 provides the requirements for responding to a seismic event during the operating phase of the life cycle of a commercial nuclear plant and is equivalent to the requirements in paragraph IV(a)(3) of appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to part 50.

Part 53 includes provisions to address staffing, training, personnel qualifications, and human factors engineering (HFE) in a manner that is risk-informed, technology-inclusive, performance-based, and flexible in nature. During the development of part 53, the staff prepared a draft white paper on “Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors,” to support interactions with stakeholders and the Advisory Committee on Reactor Safeguards (ACRS). Key considerations include the recognition that staffing, operator qualifications, and HFE are interconnected areas that must be approached in an integrated manner and, furthermore, that safety functions, including the means by which they are fulfilled, provide an effective method for informing technology-inclusive requirements.

The requirements associated with this approach are in Sec. Sec. 53.725 through 53.830. Section 53.725 discusses applicability and defines specific terms. Some definitions draw from those in Sec. 55.4. Several new definitions are introduced for use within the context of subpart F. These new definitions are the following: “Automation,” “Auxiliary operator,” “Generally licensed reactor operator,” “Interaction-dependent-mitigation facility,” “Load following,” and “Self-reliant-mitigation facility.”

Sections 53.725 through 53.830 are divided into four portions that cover general operational requirements, operator and senior operator licensing requirements, GLRO requirements, and general training requirements for plant staff. The NRC intends to provide guidance addressing the review of operator staffing plans; the review of operator, senior operator, and GLRO examination programs; and the implementation of scalable HFE reviews. Licensees will be required to use GLROs upon demonstrating compliance with the criteria in Sec. 53.800.

Certain routine communications are necessary to facilitate the operator licensing process. The NRC adapts the requirements of Sec. Sec. 55.5 and 50.74 to Sec. 53.726 to accomplish this.

Specific information must be collected in order to facilitate the initial issuance of operator licenses, as well as to allow for license renewals and required updates thereafter. Such information collection activities must also be approved by the OMB. The NRC adapts the requirements of Sec. 55.8, to include any needed updates in OMB approval information, to Sec. 53.120 to accomplish this.

The information used within the regulatory processes of the NRC must be free from omissions and inaccuracies to facilitate effective regulation. Consistent with this, the NRC adapts the requirements of Sec. 55.9 to Sec. 53.728 to require the completeness and accuracy of material information provided by individual applicants and license holders.

Section 53.730 provides performance-based and technology-inclusive requirements for assessing the role of personnel in facility safety, applying human system considerations within facility design, and incorporating operational approaches that are consistent with design- specific safety considerations. Most of these requirements are adapted from portions of Sec. Sec. 50.34(f) and 50.54 and 10 CFR part 55, “Operators' Licenses,” with considerable modification in order to reflect the introduction of new technologies and possible changes in the roles of personnel in preventing and mitigating events. The NRC intends that these technical requirements will, together, serve as a component of the required content of applications for OLs and COLs under part 53. Additionally, the NRC intends that the specific technical requirements associated with HFE, human-system interface design, concept of operations, functional requirements analysis, and function allocation will serve as a component of the required content of applications for standard DCs, standard design approvals, MLs, and CPs, as well.

Human factors engineering is essential to facilitate the role of personnel in facility safety in a manner that is both effective and reliable. The

NRC adapts Sec. 53.730(a) from the HFE design requirements of Sec. 50.34(f)(2)(iii). A key difference is that the requirement is now focused on settings where personnel fulfill their safety or emergency response roles wherever they may occur. The NRC additionally includes within the scope of this requirement activities for assuring the continued availability of plant equipment that is needed for safety, and the NRC envisions that these activities may encompass relevant maintenance, inspections, and testing as well. This requirement is associated with the staff guidance for conducting scalable reviews of HFE in DRO-ISG-2023-03, “Development of Scalable HFE Review Plans” that accompanies part 53.

Human-system interfaces provide vital information to operators across a spectrum of operating conditions that can range from normal operations through severe accident conditions. The specific types of information that must be available to support operations staff during such conditions include, in part, those associated with safety function parameters, safety system status, possible core damage states, barrier integrity, and radioactive leakage. Due to the importance of such information, the NRC requires under Sec. 53.730(b) such human-system interface design features for all facilities, irrespective of other flexibilities under part 53. Therefore, the NRC adapts specific post- Three Mile Island requirements of Sec. 50.34(f) in a technology- inclusive manner as detailed in the following:

Paragraph (b)(1) is adapted from Sec. 50.34(f)(2)(iv).

Paragraph (b)(2) is adapted from Sec. 50.34(f)(2)(v).

Paragraph (b)(3) is adapted from Sec. 50.34(f)(2)(xi), 50.34(f)(2)(xii), and 50.34(f)(2)(xxi).

Paragraph (b)(4) is adapted from Sec. 50.34(f)(2)(xvii), 50.34(f)(2)(xviii), 50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).

Paragraph (b)(5) is adapted from Sec. 50.34(f)(2)(xxvi).

Paragraph (b)(6) is adapted from Sec. 50.34(f)(2)(xxvii).

In addition to the requirements of Sec. 53.730(b)(1) through (6), a further set of human-system interface design requirements applicable only to those facilities that will be staffed by GLROs is provided under Sec. 53.730(b)(7). This prescriptive set of design requirements for those facilities that demonstrate compliance with the criteria of Sec. 53.800 recognizes that the application of HFE under Sec. 53.730(a) is anticipated to be significantly streamlined at such facilities in the absence of an expected operator role for the fulfillment of safety functions. However, it should be noted that the capability for an immediately initiated, manual reactor shutdown is conservatively mandated irrespective of any other design considerations for both interaction-dependent and self-reliant mitigation facilities, as required under Sec. 53.730(b)(8).

The NRC requires under Sec. 53.730(c) the submittal of a concept of operations that is of sufficient scope and detail to appropriately inform the staff. The development of a concept of operations can facilitate a clear understanding on the part of the NRC for potential novel operating concepts. Additionally, such information is likely to reduce the degree of resources and interactions needed for the NRC to obtain the understanding necessary to enable flexible requirements in areas such as staffing, operator qualifications, and HFE.

The NRC requires under Sec. 53.730(d) the submittal of both a Functional Requirements Analysis and a Function Allocation. The identification of design-specific safety functions and how they are fulfilled serves as a primary means for achieving technology-inclusive requirements within areas such as staffing, operator qualifications, and HFE. The Functional Requirements Analysis and Function Allocation processes (which are both HFE methods derived from systems engineering principles), provide an effective means to identify both how safety functions will be satisfied and how to characterize any associated operator role in doing so. A Functional Requirements Analysis shows what features, systems, and human actions are relied upon to demonstrate safety (i.e., fulfill safety functions). A Function Allocation then describes how safety functions are assigned to both personnel and automatic systems. However, an important adaptation of the Function Allocation for use under this final rule is the further need not only to describe allocations of safety functions to human action and automation, but also to identify allocations made to active safety features, passive safety features, or inherent safety characteristics as well.

Operating experience provides an important source of information by which to inform various aspects of facility design and operations. Accordingly, the NRC adapts in Sec. 53.730(e) the requirements of Sec. 50.34(f)(3)(i) for requiring an operating experience program.

New technologies may involve concepts of operations that are more conducive to customizable licensed operator staffing requirements than the prescriptive requirements of Sec. 50.54(m). Analyses and assessments that are based on HFE principles provide a performance- based means of determining licensed operator and senior operator staffing needed to support safe operations. In contrast, for those facilities required to be staffed by GLROs, the NRC anticipates that the operator staffing plans will reflect a simpler approach of showing that a continuity of responsibility will be maintained for facility operations throughout the operating phase, with at least one GLRO providing continuous oversight and remaining immediately available when any units are fueled. Additionally, a revised approach to the traditional position of the shift technical advisor that focuses on the availability of engineering expertise as a means of addressing uncertainties and abnormal circumstances is more suitable within the context of part 53 and is intended to be applicable to all facilities, irrespective of other design and staffing considerations.

Consistent with this approach, the NRC requires under Sec. 53.730(f) the submittal of a staffing plan that details operations staffing, how engineering expertise will be provided, and what staffing will be available to provide other needed support functions. The staffing plan description of how engineering expertise will be provided should include details of the position, such as location, expected response time, access to plant status information, and methods of communication. The staffing plan description should contain information on how the described response time has been or will be determined to be adequate based on the facility design. This requirement is associated with the staff guidance for reviewing operations staffing plans in DRO- ISG-2023-02, “Interim Staff Guidance Augmenting NUREG-1791, `Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR part 53” that accompanies part 53. Following NRC approval of the OL or COL, the staffing plan will become a condition of the facility license.

The NRC intends that, at a minimum, the approved licensed operator and senior operator (or, if applicable, GLRO) staffing, positions, and personnel locations will be incorporated into corresponding requirements within the facility TS and that a license amendment would therefore be required for any subsequent changes.

Operator training and qualification programs provide an essential

component of supporting human performance in implementing tasks with safety implications. Such programs must include components that cover the stages of initial training, examination, and continuing training. Additionally, recognizing the potential for varying concepts of operations to affect traditional, prescriptive approaches to operator proficiency, under part 53 the NRC allows facilities to develop operator proficiency programs based on facility-specific considerations.

Therefore, the NRC requires in Sec. 53.730(g)(1), as part of its approval of the OL or COL, approval of the programs that will be used for the initial training, initial examination, requalification training and examination, and proficiency of both licensed operators and senior operators. In a corresponding manner, the NRC requires in Sec. 53.730(g)(2) approval of the programs that will be used for the GLRO equivalents of each of these programs for facilities with such staffing. The NRC intends that examination program requirements will be associated with staff guidance for the review of tailored examination processes that are planned to accompany part 53. Following the completion of an initial training program, continuing training programs provide an important means of sustaining the knowledge and abilities of individuals. The NRC adapts the requirements of Sec. 50.54(i-1) in Sec. 53.730(g)(3) to require that operator continuing training programs be in effect to support operator performance. Under part 53, the NRC requires these programs to be in effect concurrent with when the initial operator examinations first commence, in effect putting the programs in place only when they are needed. This represents a modification of the comparable requirement of Sec. 50.54(i-1), which links the commencement of these programs to a timeline driven by the licensing of the facility.

The authorization to manipulate controls of the facility that directly affect reactivity or power level is restricted to individuals who are either licensed operators, licensed senior operators, or GLROs. However, for practical purposes, situations in which an individual is participating in an approved training program or reestablishing proficiency may also call for them to operate the controls of the facility under the cognizance of a licensed individual. The NRC adapts the requirements of Sec. 55.13 in Sec. 53.735 to accomplish this, with a notable difference being the incorporation of GLROs.

Section 53.740 provides requirements for OL and COL holders under part 53. Portions of Sec. 53.740 are adapted from the conditions of Sec. 50.54. In general, the conditions for operations staffing under part 53 reflect considerations for potential technological differences and varying concepts of operation that are expected among part 53 facility licensees. Additionally, certain requirements are specific to the operating phase while others remain in effect following the permanent cessation of facility operations during the decommissioning phase.

All commercial nuclear plants licensed under part 53 require some form of licensed operator staffing, whether it be by specifically or generally licensed operators. Consistent with this, the NRC requires under Sec. 53.740(a) that facility licensees demonstrate compliance with the programmatic requirements for either specifically licensed operators and senior operators or for GLROs, as applicable to the facility.

The NRC recognizes that technology-inclusive facility staffing will need to account for a potentially wide range of concepts of operations; for this reason, flexible and performance-based approaches for establishing required facility staffing are appropriate. However, once the appropriate facility staffing has been determined and approved by the NRC, such staffing must be maintained to ensure that the appropriately qualified individuals will be available when needed to support the safe operation of the facility. Therefore, the NRC requires under Sec. 53.740(b) that the staffing described within the approved facility staffing plan be maintained as a condition of the facility license as opposed to prescriptive staffing requirements like those of Sec. 50.54(k) and (m).

Because operation of facility controls directly affects reactivity or power level, only those individuals who possess appropriate levels of qualification and authorization are permitted to operate those controls. The NRC adapts the requirements of Sec. 50.54(i) in Sec. 53.740(c) to require that only specifically licensed operators and senior operators or, alternatively, GLROs, may operate facility controls, with allowance for specified exceptions for the purposes of operator training or proficiency.

Senior operators, by virtue of their license level, are qualified and authorized both to perform certain important responsibilities and to direct the licensed activities of licensed operators. Therefore, facilities that are required to be staffed by specifically licensed operators must also include senior operators within their staffing. In contrast, facilities staffed with GLROs only have a single license level available and, therefore, there is no equivalent provision for such facilities. The NRC adapts the requirements of Sec. 50.54(l) in Sec. 53.740(d) to require the licensing and designation of senior operators at facilities staffed by specifically licensed operators.

In contrast with control manipulations that directly affect reactor power and reactivity (e.g., control rod movement, control drum rotation, recirculation pump speed adjustment, reactor coolant system boration or dilution, etc.) and are therefore restricted to performance only by licensed operators, other types of plant operations that may result in reactor power and reactivity changes via means that are indirect in nature (e.g., electrical generation changes, turbine bypass valve operation, steam usage by process heat applications, etc.) may be implemented by non-licensed personnel. However, due to the potential influence of such operations on reactor power and reactivity, the continuous oversight of reactor parameters by a licensed operator is necessary during these operations. The NRC therefore adapts the requirements of Sec. 50.54(j) in Sec. 53.740(e) to require appropriate oversight of operations, other than those associated with the controls themselves, that may affect reactivity or power level.

Load following where plant output automatically changes in response to externally originated instructions or signals is not permitted under the existing regulations of Sec. 50.54. However, new technological considerations and concepts of operation may justify such an operational approach under appropriate circumstances. The NRC recognizes that, beyond electrical power generation, load following may also affect other applications of plant output, such as hydrogen production, desalination, or district heating. For load following to be permissible, measures must be in place to provide assurance that plant output considerations are not permitted to lead to challenges to safe reactor operations. These measures may consist of automated control systems, automatic protective features, or the continuous oversight and immediate intervention capability of an appropriately qualified and authorized individual. Section 53.740(f) allows for load following, provided that appropriate measures are in place. In considering the acceptability of the measures associated with load following, the NRC expects that any automatic protection relied

upon would be separate from that credited for reactor protection purposes and would employ setpoints that are set so as to prevent actuation of the reactor protection system while accomplishing its functions to the extent practical.

Core alterations such as refueling are associated with specific considerations that warrant limiting the oversight of such operations to appropriately qualified and authorized individuals. Unlike other types of fuel handling operations, core alterations occur within the confines of a reactor vessel that is specifically designed to support and sustain nuclear criticality, thereby justifying the imposition of higher qualification levels within such contexts. The NRC adapts the requirements of Sec. 50.54(m)(2)(iv) in Sec. 53.740(g) to require the supervision of core alterations by either a specifically licensed senior operator, a specifically licensed senior operator whose license is limited to fuel handling, or by a GLRO, as applicable to the facility. Because certain commercial reactor designs may be capable of refueling while at power and, in any event, overall facility oversight will already be required by either a specifically licensed senior operator or by a GLRO, the NRC omits this requirement as redundant during periods where core alterations occur while the plant is operating.

It is impossible to predict every possible scenario that a commercial nuclear plant might potentially encounter. Therefore, it is prudent to grant the authority for appropriately qualified individuals to depart from facility license conditions when emergency circumstances dictate that doing so is in the interest of public health and safety. The NRC adapts the requirements of Sec. 50.54(x) and (y) in Sec. 53.740(h) to permit specific individuals to authorize departures from facility license conditions or TSs when emergency conditions warrant doing so for the protection of the public health and safety. Recognizing that certain facilities licensed under part 53 may be staffed by GLROs in lieu of specifically licensed senior operators, the NRC extends this authority to GLROs. While it is not anticipated that GLROs will have a role in the fulfillment of safety functions at self- reliant-mitigation facilities, nor is it anticipated that operators at such facilities would be in a position by which to significantly influence radiological safety outcomes, the very nature of the Sec. 50.54(x) and (y) and Sec. 53.740(h) provisions concern situations that are unanticipated and, therefore, unforeseeable. Thus, it is appropriate to grant GLROs a comparable authority to that of senior licensed operators and certified fuel handlers as it relates to invoking this provision under emergency conditions as a means of accounting for such possibilities.

Due to the unique authorities and responsibilities of both specifically and generally licensed reactor operators, it is essential that any individual fulfilling such a role demonstrate compliance with the regulatory requirements for operator licensing. Section 107 of the AEA authorizes the Commission to prescribe conditions for the licensing of operators and to issue licenses consistent with those conditions. The NRC adapts the requirements of Sec. 55.3 in Sec. 53.745 to require that any person performing the function of an operator, senior operator, or GLRO must be authorized by a license issued by the Commission.

The NRC will license individuals as operators under both specific and general licensing frameworks. Specific licenses will be for licensed operators (i.e., reactor operators) and senior operators (i.e., senior reactor operators) and will be issued to a named person upon approval by the Commission of an application for that named person. In contrast, GLROs will perform duties under the provisions of a general license that is effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. The NRC sets forth requirements for the use of a specific licensing process for licensed operators and senior operators under Sec. Sec. 53.760 through 53.795, with Sec. 53.760 addressing applicability.

Medical fitness is an important component of the overall process of specifically licensing operators because it provides assurance that operators will be able to carry out important duties without being precluded from doing so by health-related issues. Medical fitness also provides assurance that such issues will not adversely affect the performance of assigned job duties or cause operational errors that endanger public health and safety. In addition to a requirement for medical fitness, a medical examination by a physician to confirm compliance with this requirement is necessary. The NRC adapts the requirements of Sec. Sec. 55.21, 55.23, and 55.27 under Sec. 53.765 to require medical fitness, examinations by physicians, and medical certification for specifically licensed operators and senior operators. In recognition of the fact that GLROs are not expected to have a role in the fulfillment of safety functions at the facilities at which they are licensed, the NRC does not extend a comparable medical requirement to GLROs.

The NRC also adapts the requirements of Sec. Sec. 55.25 and 50.74(c) in Sec. 53.770 to require that timely notifications be made to the NRC if a specifically licensed operator or senior operator develops a permanent physical or mental condition that adversely affects the performance of assigned operator job duties or could cause operational errors endangering public health and safety. Notwithstanding this requirement related to permanent medical conditions, the NRC continues to recognize that it is appropriate for facility licenses to impose administrative restrictions and conditions upon specifically licensed operators and senior operators in response to temporary medical conditions.

The process of specifically licensing individuals as licensed operators or senior operators requires the submittal of applications to the NRC for review. These applications must detail certain elements associated with licensing, including the demonstration of compliance with examination, experience, and medical requirements. The NRC adapts the requirements of Sec. Sec. 55.31 through 55.35 in Sec. 53.775 to include requirements for the applications associated with the specific licensing of licensed operators and senior operators at commercial nuclear plants licensed under part 53. In contrast with the part 55 requirements, the NRC provides additional flexibility by locating certain details associated with the preparation and submittal of these applications within guidance in lieu of placement within this final rule itself.

The NRC includes overall programmatic requirements for specifically licensed operator and senior operator training, examination, and proficiency in Sec. 53.780. In general, the requirements are adapted from those in part 55, with several additional flexibilities being incorporated to better account for potential variations in reactor technologies and concepts of operations. The requirements in Sec. 53.780 cover, in part, the initial training, initial examination, requalification training, requalification examination, and proficiency of specifically licensed operators and senior operators.

The initial training process provides individuals with the knowledge and abilities needed to subsequently fulfill assigned duties as licensed operators or senior operators in a safe and reliable manner. The use of a systems approach to training (SAT) ensures that the

training program is based upon job requirements in a manner that can be adapted to account for differences in plant technology, concepts of operations, and operator roles in the fulfillment of design-specific safety functions. The NRC requires under Sec. 53.780(a) that facility licensees implement a SAT-based training program for the initial training of licensed operator and senior operator applicants. The program must be adequate to ensure that applicants will be capable of performing the duties necessary both to protect public health and safety and to maintain plant safety functions. The NRC further requires that such programs be subject to NRC approval and subsequent change control processes of an appropriate nature.

Examinations provide a means of assessing that individuals have achieved a degree of knowledge and ability that is sufficient to carry out assigned duties as licensed operators or senior operators in a manner that is safe and reliable. The NRC adapts the requirements of Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec. 53.780(b) to require that facilities establish and implement an initial examination program. However, a key difference from the comparable requirements of part 55 is that facilities have the flexibility to propose, subject to NRC approval, the examination methods and criteria to be used in assessing satisfactory applicant performance. Such examination programs (including those used within the scope of requalification training) must provide for acceptable levels of both test validity and test reliability in order to be considered acceptable. The NRC intends that staff guidance will be available to facilitate the review of licensing examination programs that are proposed by facility licensees and that, following NRC approval, initial examination programs will be subject to an appropriate change control process. Furthermore, the NRC provides holders of licenses to operate commercial nuclear plants under part 53 the alternative of administering their own approved licensing examinations. The NRC will continue to exercise appropriate oversight of the program, make operator licensing decisions based upon the examination results, and reserve the right to administer the examinations in lieu of permitting the facility to do so. However, irrespective of the provided flexibilities in examination format and structure, at a minimum, topics from the following general categories of knowledge and abilities should be sampled in such examinations:

Reactor Theory, Thermodynamics, and Chemical Interactions

Plant Systems and Components

Reactivity Management and Manipulations

Radiation Control and Safety

Emergency, Abnormal, and Normal Operations

Administrative Requirements and Conditions of the Facility License

Requalification training programs provide for the continuing training and examination of specifically licensed operators and senior operators to ensure that they maintain the knowledge and abilities needed to support the safe and reliable performance of job duties following the completion of an initial training and examination program. The NRC adapts the requirements of Sec. 55.59 in Sec. 53.780(c) to require that facilities implement both a SAT-based requalification training program and a biennial requalification examination program. However, a notable difference from the biennial requalification examinations required under part 55 is that distinct annual operating test and biennial written examination components are not mandated, with the facility licensee instead proposing the examination methods and criteria to be used in assessing satisfactory performance. The NRC intends that guidance will be available to facilitate the review of the requalification examination programs that are proposed by facility licensees and that, following NRC approval, requalification examination programs will be subject to an appropriate change control process.

For examinations to provide valid assessments of the knowledge and abilities of individuals, the examinations must remain free from compromises that could affect their underlying integrity. The NRC adapts the requirements of Sec. 55.49 in Sec. 53.780(d) to require that examinations and related activities remain free from any compromise that might affect the integrity of the examination process.

Simulators provide a valuable means of training and evaluating plant operators, and the NRC is specifically authorized under the Nuclear Waste Policy Act of 1982, as amended (NWPA), section 306 (42 U.S.C. 10226) to establish regulations for the use of simulators within such context. The NRC adapts the requirements of Sec. 55.46 in Sec. 53.780(e) to address the use of simulation facilities for training, examinations, and applicant experience requirements, as well as to address the maintenance of simulator fidelity. However, the requirements of part 53 do not mandate that full scope, plant- referenced simulators be used and will allow the use of alternative simulation facilities consisting of, for example, partial scope simulators or the plant itself, provided that all associated requirements can be demonstrated to be met using alternative approaches and methods. Additionally, in allowing for the possibility that an applicant or licensee might demonstrate compliance with training, examination, or experience requirements using the plant itself, the NRC is not allowing the initiation of transients on the actual plant. Consistent with this, aside from controlled reactivity manipulations that are conducted for the purposes of demonstrating compliance with experience requirements, actual plant components may not be operated for these purposes. Rather, the NRC perspective is that the use of the plant for training and examination purposes should be restricted to techniques such as walkthroughs, job performance measures, simulated tasks, use of augmented reality technology, and similar approaches that provide training and examination value while avoiding the operation of actual plant components.

There may be situations in which applicants for operator or senior operator licenses have previous training and experience that justifies waiving some, or all, of the initial examination requirements. The NRC adapts the requirements of Sec. 55.47 in Sec. 53.780(f) to allow for consideration of requests for waivers of examinations requirements. In contrast with the part 55 requirements, the NRC locates certain details associated with such waiver requests within guidance documentation in lieu of placement within this final rule itself.

For licensed operators and senior operators to perform their assigned duties safely and reliably, it is essential that they perform those duties frequently enough so as to maintain a sufficient degree of proficiency. The NRC adapts the requirements of Sec. 55.53(e) and (f) in Sec. 53.780(g) to require that specifically licensed operators and senior operators maintain proficiency and, if proficiency is not maintained, regain proficiency prior to resuming licensed duties. However, in recognition of the fact that varying concepts of operations are possible for advanced reactor facilities, the NRC, in contrast with the requirements of part 55, is allowing facility licensees to establish their own programs for operator proficiency, subject to NRC approval.

As the holders of specific licenses, licensed operators and senior operators

must be subject to license conditions on an individual basis to ensure that the basis upon which the licenses were issued remains valid. The NRC adapts the requirements of Sec. 55.53 in Sec. 53.785 to require appropriate conditions of licenses for specifically licensed operators and senior operators. However, in contrast with the requirements of Sec. 55.53(e) and (f), the NRC is allowing certain aspects of operator proficiency to be addressed by an NRC-approved facility proficiency program.

Licenses for specifically licensed operators and senior operators are issued by the NRC and must remain subject to modification or revocation. The NRC adapts the requirements of Sec. Sec. 55.51 and 55.61 in Sec. 53.790 to address the issuance, modification, and revocation of licenses issued to specifically licensed operators and senior operators.

The licenses issued to specifically licensed operators and senior operators are valid for a period of 6 years, after which they expire, unless otherwise renewed. The NRC adapts the requirements of Sec. Sec. 55.55 and 55.57 in Sec. 53.795 to address the expiration and renewal of licenses issued to specifically licensed operators and senior operators.

In developing this final rule, the NRC has discussed with stakeholders the considerations that might justify the omission of the specifically licensed operators and senior operators. However, even for an inherently safe reactor with autonomous operation features, certain important administrative functions (e.g., compliance with TS, operability determinations, NRC notifications, emergency declarations, risk assessment, maintenance oversight, and radiological release limit compliance) would still need to be accomplished by appropriately qualified and authorized individuals. Additionally, the NRC recognized that manual manipulations of facility reactivity controls must only be performed by individuals who have been appropriately licensed by the Commission. The NRC therefore establishes under Sec. 53.800 a new class of facility (defined as a self-reliant-mitigation facility), according to the criteria contained in Sec. 53.800 for part 53. These facilities will employ GLROs rather than specifically licensed operators and senior operators. The GLRO regulations offer enhanced flexibilities and targeted relaxations in a manner that is commensurate with the modified role of such operators to ensure the safe operation of the associated facilities. In contrast, those facilities not meeting the criteria of Sec. 53.800 will instead be considered interaction- dependent-mitigation facilities and will require staffing by specifically licensed operators and senior operators. The terminology used to designate these facility types reflects differences in how operators are anticipated to need to interact with their plant systems in mitigating events and achieving safe outcomes; such systems may either need operators to interact with them in some manner (i.e., be interaction-dependent) or may instead be able to rely fully upon their own capabilities independent of operator interaction (i.e., be self- reliant).

Generally licensed reactor operators differ from specifically licensed operators because the latter will be directly and independently evaluated by the NRC as part of their licensing process. This direct and independent evaluation remains appropriate when operators may reasonably be expected to exert a significant influence on public health and safety outcomes. Therefore, a key determinant as to whether generally licensed reactor operators can be utilized in facility staffing is the assessment of the operator's role in maintaining and fulfilling safety functions at the facility, such as through the performance of credited actions for the mitigation of plant events.

The criteria in Sec. 53.800 designate self-reliant-mitigation facilities. These criteria are derived from the following set of considerations:

no human action needed to satisfy radiological consequence criteria;

no human action needed to address LBEs;

safety functions not allocated to human action;

reliance upon robust and highly reliable safety features; and

appropriate defense in depth achieved without reliance on important human action.

It should be noted that those facilities not meeting the criteria in Sec. 53.800 will instead be classified as interaction-dependent- mitigation facilities and will require staffing by specifically licensed operators and senior operators instead.

Generally licensed reactor operators will perform duties under the provisions of a general license that is effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. The NRC sets forth requirements for the general licensing process for GLROs under Sec. Sec. 53.805 through 53.820. The requirements for GLROs parallel those for senior operators in regard to their comparable administrative responsibilities. Nonetheless, the requirements for GLROs are relaxed and incorporate greater flexibilities compared to the requirements for specifically licensed operators in a manner that is consistent with the GLRO's role in safety at self-reliant-mitigation facilities.

In order to use GLROs in lieu of specifically licensed operators and senior operators, a OL/COL applicant must demonstrate that its proposed facility is a self-reliant-mitigation facility, i.e., that it will comply with the following requirements on an ongoing basis: maintaining GLRO qualifications for the performance of important functions and tasks; incorporating relevant programmatic controls into TS; administering the related programs for training, examination, and proficiency; and ensuring that the relevant provisions of parts 26 and 73 are met. Additionally, to provide for an accurate accounting of what individuals are licensed under the general license, facility licensees are required to report the identities of all generally licensed reactor operators to the NRC on an annual basis. Furthermore, a facility licensee must ensure that the facility design and performance continue to meet the technological criteria to be classified as a self-reliant- mitigation facility (i.e., the criteria of Sec. 53.800) on a continual basis during the operating phase, as the relaxations afforded to such facilities in the areas of operator licensing, staffing, and HFE are predicated on this assumption. The NRC therefore establishes under Sec. 53.805 requirements for facility licensees that address issues such as these. Finally, the failure of a self-reliant-mitigation facility to subsequently meet the criteria of Sec. 53.800 after the issuance of an OL or COL will constitute a reportable event (i.e., an unanalyzed condition that significantly degrades plant safety) under the provisions of Sec. 53.1630.

The NRC sets forth the general license for GLROs under Sec. 53.810. GLROs will be licensed as a class of individuals under the provision of Sec. 53.810(a) and will be subject to the conditions specified in Sec. 53.810(b) through (g). Portions of these conditions are adapted from Sec. 55.53 and from those conditions currently included in the licenses issued to specifically licensed operators and senior operators. The NRC retains the ability to suspend or prohibit individuals from operating under the general license should such action be warranted.

The NRC includes overall programmatic requirements for GLRO training, examination, and proficiency under Sec. 53.815. In general, these

requirements are adapted from those of part 55 and parallel those also included for specifically licensed senior operators in Sec. 53.780. These requirements include increased flexibilities and several targeted relaxations that reflect the limited role of GLROs in facility safety. The requirements under Sec. 53.815 cover, in part, the initial training, initial examination, continuing training, requalification examination, and proficiency of GLROs. Section 53.805 requires the facility licensee to develop, implement, and maintain these programs. Section 53.810, in turn, prescribes that the requirements of Sec. 53.805 must be met as a requirement of the general license. The implication of this structure is that the facility licensee must implement these programs for training, examination, and proficiency, and GLROs must participate in these programs to demonstrate compliance with the requirements of the general license.

The initial training process provides GLROs with the knowledge and abilities needed to fulfill assigned duties as GLROs. The use of an SAT serves to ensure that the training program is based upon job requirements in a manner that can be adapted to account for differences in plant technology and concepts of operations. The NRC requires under Sec. 53.815(b) that facility licensees implement a SAT-based training program for the initial training of GLROs that is adequate to ensure that they have the necessary knowledge, skills, and abilities to perform their duties. The NRC further requires that such programs be subject to NRC approval, oversight, and appropriate change control processes. The training program must ensure that GLROs maintain the necessary knowledge, skills, and abilities.

Examinations provide a means of assessing that individuals have achieved a degree of knowledge and ability that will be sufficient to enable them to carry out assigned duties as GLROs in a manner that is both safe and reliable. The NRC adapts the requirements of Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec. 53.815(b) to require that facility licensees establish and implement an initial examination program. A key difference from the comparable requirements of part 55 is that facility licensees are afforded the flexibility to propose, subject to NRC approval, the examination methods and criteria to be used in assessing satisfactory individual performance. Such examination programs (including those used within the scope of continuing training) must provide for acceptable levels of both test validity and test reliability in order to be considered acceptable. The NRC intends that staff guidance will be available to facilitate the review of initial examination programs that are proposed by facility licensees and that approved initial examination programs will be subject to an appropriate change control process. In contrast with both the requirements of part 55 and the requirements of Sec. 53.780, the NRC does not intend to administer or evaluate these initial examinations. However, the examination processes themselves will continue to be subject to ongoing NRC oversight. Irrespective of the provided flexibilities in examination format and structure, topics from the following general categories of knowledge and abilities should be sampled in such examinations:

Reactor Theory, Thermodynamics, and Chemical Interactions

Plant Systems and Components

Reactivity Management and Manipulations

Radiation Control and Safety

Emergency, Abnormal, and Normal Operations

Administrative Requirements and Conditions of the Facility License

Continuing training programs provide the ongoing training and examination of GLROs to ensure that they maintain the knowledge and abilities needed to support the safe and reliable performance of job duties following the completion of an initial training and examination program. The NRC adapts the requirements of Sec. 55.59 in Sec. 53.815(b) to require that facility licensees implement both an SAT- based continuing training program and a requalification examination program. However, a notable difference from the examinations required under part 55 is that distinct annual operating test and biennial written examination components are not mandated. The facility licensee will instead propose examination methods and criteria to be used in assessing satisfactory performance. Furthermore, unlike the comparable requirements of part 55 and those for specifically licensed operators and senior operators, a biennial periodicity for requalification examinations is not prescribed. However, adequate justification for the proposed periodicity of requalification examinations is required. The NRC intends that staff guidance will be available to facilitate the review of the requalification examination programs that are proposed by facility licensees. Approved requalification examination programs will be subject to an appropriate change control process.

For examinations to provide for valid assessments of the knowledge and abilities of individuals, the examinations must remain free from compromises that could affect their underlying integrity. The NRC adapts the requirements of Sec. 55.49 in Sec. 53.815(d) to require that examinations and related activities remain free from any compromise that might affect the integrity of the examination process.

Simulators provide a valuable means of training and evaluating plant operators and the NRC is specifically authorized under the NWPA, section 306 (42 U.S.C. 10226) to establish regulations for the use of simulators within such context. The NRC adapts the requirements of Sec. 55.46 in Sec. 53.815(e) to address the use of simulation facilities for training and examinations, and experience requirements, as well as to address the maintenance of simulator fidelity. The use of full scope, plant-referenced simulators is not mandated. The potential use of alternative simulation facilities consisting of, for example, partial scope simulators or the plant itself, is allowed provided that all associated requirements are demonstrated to be met using alternative approaches and methods. Additionally, in allowing for the possibility that an applicant or licensee might demonstrate compliance with training and examination requirements using the plant itself, the NRC is not allowing the initiation of transients on the actual plant. Consistent with this, aside from controlled reactivity manipulations that are conducted for the purposes of demonstrating compliance with experience requirements, actual plant components may not be operated for these purposes. Rather, the use of the plant for training and examination purposes should be restricted to techniques such as walkthroughs, job performance measures, simulated tasks, use of augmented reality technology, and similar approaches that provide training and examination value while avoiding the operation of actual plant components.

There may be situations in which GLROs have previous training and experience that justifies waiving some, or all, of the initial examination. Therefore, under Sec. 53.815(f) the NRC allows facility licensees to waive some, or all, portions of initial examinations provided that such waivers are consistent with a program that has been approved by the NRC.

For GLROs to safely and reliably perform their assigned duties, it is essential that they perform those duties frequently enough so as to maintain a sufficient degree of proficiency.

However, the NRC recognizes that facilities that utilize GLROs may have concepts of operation that warrant unique proficiency considerations. Therefore, the NRC requires in Sec. 53.815(g) that facility licensees develop, implement, and maintain programs to maintain and reestablish, if needed, the proficiency of GLROs. This could occur, for example, if an individual's extended absence from watch standing has rendered proficiency requirements unmet.

The general license should remain in effect for an individual only while that individual remains employed in a position that may call for the individual to manipulate the reactivity controls of the facility. The NRC requires under Sec. 53.820 that the general license ceases to be applicable on an individual basis when an individual's employment status becomes such that this is no longer the case. However, the NRC recognizes that for some types of self-reliant-mitigation facilities, very long periods may elapse between circumstances that necessitate manual manipulation of reactivity controls. Therefore, the general license remains in effect for an individual as long as the individual's current position could potentially require that individual to manipulate reactivity controls at some point within the course of the individual's assigned job duties.

The NWPA, section 306 (42 U.S.C. 10226) authorizes and directs the NRC to, in part, issue regulations and guidance that address the training and qualifications of civilian nuclear power plant operators, supervisors, technicians, and other appropriate operating personnel. The NRC implements this in part 50 through the requirements of Sec. 50.120, “Training and qualification of nuclear power plant personnel.” The NRC adapts under Sec. 53.830, with modifications, the requirements of Sec. 50.120 for use in part 53 to provide more flexible personnel training and qualification requirements than those in Sec. 50.120 and better reflect diverse concepts of operations.

The NRC recognizes that the categories of nuclear power plant personnel in Sec. 50.120 may not be needed for the diverse concepts of operations, staffing models, and non-traditional personnel roles and responsibilities anticipated under part 53; conversely, and for the same reasons, additional categories of personnel may need to be covered by part 53. The NRC also recognizes that the timeframe prescribed in Sec. 50.120 for the establishment of training programs may not be aligned with the schedules associated with the startup of certain types of commercial nuclear plant facilities. However, the NRC also recognizes that the SAT-based training required under Sec. 50.120 remains an appropriate means by which training programs should continue to be developed and implemented. Therefore, the approach taken by the NRC in addressing the training of certain plant staff under part 53 reflects greater flexibilities in personnel categories and programmatic timeframes, while still retaining the requirement that such training programs be based on SAT.

The NRC requires under Sec. 53.830 SAT-based training programs with the timeframe for when such programs are required being based upon when the associated personnel are needed to support facility-specific needs. The training programs will cover the training and qualification of personnel in the general categories of supervisors, technicians, and other appropriate operating personnel. Regarding the category of supervisors, this is intended to reflect on-shift supervisors for the licensed operators, similar to the current classification in Sec. 50.120(b)(2)(iii), but Sec. 53.830 uses language that is less specific to account for different conduct of operations and organizational structures for commercial nuclear plants which may require greater regulatory flexibility. The licensee is not required to seek NRC approval of a training program prior to usage. However, the licensee is required to accommodate NRC inspection of the training program. The NRC intends to develop guidance to facilitate the inspection of these training programs but does not intend for such guidance to preclude the potential for the training programs to be maintained by a separate, NRC-approved accreditation process.

Section 53.845 requires programs to be developed, implemented, and maintained to help ensure that design features and human actions have the capabilities and reliabilities necessary to demonstrate compliance with the safety criteria in subpart B throughout the operating life of each commercial nuclear plant. The programmatic requirements in subpart F also address areas such as radiation protection needed to control routine effluents during normal operations. Sections 53.850 through 53.910 require programs to support specific activities needed to ensure the prevention or mitigation of unplanned events or to support normal operations for any reactor design. However, each holder of an OL or COL is required to assess whether additional programs are needed for the specific reactor design and location of the commercial nuclear plant. Licensees are able to combine, separate, and otherwise organize programs and related documents as appropriate for the technologies and organizations associated with the commercial nuclear plant.

Section 53.850 requires a radiation protection program associated with the requirements in subparts B and C for public doses resulting from normal operations and the protection of plant workers. The requirements related to doses from normal operations, including routine effluents, are similar to those specified in Sec. 50.36a, “Technical specifications on effluents from nuclear power reactors,” and related requirements in standard TS for offsite dose calculation manuals. While the section includes requirements that are technically and programmatically similar to part 50, Sec. 53.850 does not include a requirement for effluent-related TS as is required in Sec. 50.36a. A requirement similar to that found in the administrative controls section of TS for operating reactors licensed under parts 50 and 52 is included for programmatic controls of solid wastes to complement the design requirements in Sec. 53.425.

Section 53.855 requires an emergency response plan that demonstrates compliance with the requirements in appendix E to part 50 and Sec. 50.47(b) or Sec. 50.160. The regulations in Sec. 50.47 stating that the NRC will not issue certain licenses unless it finds that there is reasonable assurance that adequate protective measures can and will be taken to protect public health and safety in the event of a radiological emergency apply equally to applications under part 53 complying with the applicable standards set forth in either Sec. 50.160 or the requirements in appendix E to part 50 and Sec. 50.47(b).

In its 2008 Advanced Reactor Policy Statement, the Commission stated their expectation that “the safety features of advanced reactor designs will be complemented by the operational program for Emergency Planning (EP). This EP operational program, in turn, must be demonstrated by inspections, tests, analyses, and acceptance criteria to ensure effective implementation of established measures.” Consistent with this policy statement, emergency plans and emergency planning zones are not safety features in the design. In SECY-97-020, “Results of Evaluation of Emergency Planning for Evolutionary and Advanced Reactors,” dated January 27, 1997, the staff indicated that the rationale upon which EP for current reactor designs is based, that is, potential consequences from a spectrum of accidents, is appropriate for use as the basis for EP for evolutionary and

passive advanced LWR designs and is consistent with the Commission's defense-in-depth safety philosophy. Also, in its Safety Goals Policy Statement the Commission stated that: “A defense-in-depth approach has been mandated in order to prevent accidents from happening and to mitigate their consequences. Siting in less populated areas is emphasized. Furthermore, emergency response capabilities are mandated to provide additional defense-in-depth protection to the surrounding population.” Consistent with this policy statement, Sec. 53.855 contributes an additional independent layer of defense in depth for commercial nuclear plants. Therefore, the emergency plans and emergency planning zones under Sec. 53.855 are not used to demonstrate compliance with subpart B and subpart C of part 53. Rather, compliance with the requirements in Sec. 53.855 provides reasonable assurance that adequate protective measures can and will be taken to protect public health and safety in the event of a radiological emergency.

Section 53.860 identifies the applicable regulations for part 53 applicants related to the programs for physical security, cybersecurity, FFD, AA, and information security. These programs are discussed in more detail in section IV, “Changes to Other Parts of 10 CFR,” of this document.

Section 53.860(a) requires licensees to develop, implement, and maintain a physical protection program that meets either Sec. 73.55 or Sec. 73.100, and includes physical protection of SNM and Category 1 and Category 2 radioactive material, if applicable.

Section 53.860(b) requires licensees to establish, implement, and maintain an FFD program under part 26. Section 53.860(c) requires licensees to establish, implement, and maintain an AA program in accordance with either Sec. 73.56 or Sec. 73.120, as appropriate. Section 53.860(d) requires licensees to establish, implement, and maintain a cybersecurity program in accordance with either Sec. 73.54 or Sec. 73.110. Section 53.860(e) requires licensees to establish, implement, and maintain an information protection system that complies with the requirements of Sec. Sec. 73.21, 73.22, and 73.23, as applicable.

Section 53.865 establishes requirements for quality assurance and refers to appendix B to part 50 for the part 53 requirements for SR design features. Requirements related to evaluating and reporting changes to the quality assurance program are included in subpart I and are equivalent to those found in Sec. 50.54.

Section 53.870 requires licensees to actively assess possible degradation of SSCs from the effects of aging, fatigue, and environmental conditions. The inclusion of requirements related to designing and monitoring for possible degradation mechanisms reflects important lessons learned from the history of LWRs and the likely introduction of new design features and materials in future commercial nuclear plants. The allowable combinations of design features, operating experience, testing, and monitoring during operations support performance-based approaches to the initial licensing of new technologies. The performance-based approach to integrity assessment programs also allows for the subsequent consideration of operating experience and appropriate corrective actions or allowable relaxations for ensuring that design features comply with the functional design criteria of Sec. Sec. 53.410 and 53.420. The program is based upon a comprehensive and integrated evaluation of the aging and other degradation mechanisms applicable to the design; identification of the affected SSCs; the allowances provided in the design of the SSCs for degradation; and schedules and procedures for determining if and at what rate degradation is occurring, as well as its cause. Risk insights can be used to prioritize the monitoring, evaluation, and management of degradation based upon the importance of the SSC to safety and the time frame for when the effects of degradation could be of concern.

Section 53.875 establishes requirements for a fire protection program supporting operations similar to Sec. 50.48. The fire protection program during operations will work in concert with specific fire protection requirements in subpart C for design and analyses and in subpart E for construction and manufacturing.

Section 53.880 establishes requirements for an inservice inspection (ISI) and inservice testing (IST) program, which are historically important activities conducted in accordance with ASME codes and regulations in Sec. 50.55a. While part 53 does not incorporate specific consensus codes and standards into the regulations, Sec. 53.880 allows for the use of generally accepted codes and standards. The requirement for an ISI and IST program reinforces the need to develop monitoring programs to be conducted during a plant's operations phase to complement the design process and address inherent uncertainties. The NRC encourages the continued use of consensus codes and standards supporting design, testing, and inspections to support integrated and performance-based approaches in demonstrating compliance with the requirements in part 53.

Section 53.910 establishes requirements for developing, implementing, and maintaining procedures (e.g., operations and emergency operating procedures) and guidelines (e.g., accident management guidelines). The programmatic requirements for many of the procedures listed in this section are similar to the requirements found in the administrative controls section of TS for plants licensed under parts 50 and 52. The inclusion, where appropriate, of accident management guidelines in these requirements is intended to ensure that an integrated set of procedures and guidelines is established by licensees to ensure command and control across the spectrum of possible event sequences. The required procedures also include those needed to complement the design requirements in Sec. 53.440(m) related to criticality alarms and the equivalent of the procedures required in Sec. 50.54(hh) to address notifications of potential aircraft threats.

Subpart G provides the regulatory requirements for the decommissioning phase of the life cycle of a commercial nuclear plant. The requirements in subpart G for the decommissioning of a commercial nuclear plant are adapted from the current regulations in Sec. 50.75, “Reporting and recordkeeping for decommissioning planning,” Sec. 50.82, “Termination of license,” and Sec. 50.83, “Release of part of a power reactor facility or site for unrestricted use.” Although the requirements from those sections of part 50 have been copied into subpart G with relatively few changes, the requirements are reorganized to fit within the part 53 structure. The few changes made were primarily to make the requirements more technology-inclusive by adding alternatives within sections, whereas some requirements in part 50 were developed specifically for LWRs.

As an example, Sec. 50.75 provides minimum amounts of decommissioning funds required to demonstrate reasonable assurance of funds for decommissioning LWRs. Such generic amounts have not been developed for all reactor technologies that may be licensed under part 53. Therefore, a requirement is included in Sec. 53.1020, “Cost estimates for decommissioning,” for site-specific cost estimates for decommissioning to be developed

considering costs in such areas as engineering, labor, and waste disposal. The derivation of the generic cost estimates for LWRs in Sec. 50.75 is provided in NUREG/CR-5884, “Revised Analyses of Decommissioning for the Reference Pressurized Water Reactor Power Station,” and NUREG/CR-6187, “Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station.” Similar to part 50, a provision for an annual adjustment of decommissioning cost estimates is included in Sec. 53.1030.

The NRC is currently pursuing another rulemaking, “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning,” which was published as a proposed rule for public comment on March 3, 2022 (87 FR 12254). As these rulemakings progress, the NRC will consider revisions to part 53 to align the two rulemaking efforts. For example, Sec. 53.1075 could be expanded to include or reference requirements for decommissioning in areas such as EP and security in addition to the decommissioning fire protection plans that provide an equivalent to Sec. 50.48(f).

Subpart H provides requirements related to applications under part 53 for NRC licenses, certifications, or approvals for commercial nuclear plants.

Subpart H specifies requirements applicable to all part 53 applications as well as requirements specific to part 53 applications for LWAs, ESPs, standard design approvals, standard DCs, MLs, CPs, OLs, and COLs. Subpart H is equivalent to and includes all existing licensing, certification, and approval processes currently covered under parts 50 and 52, with the exception of the process for early review of site suitability issues. Interactions with external stakeholders during the development of the proposed rule did not identify significant interest in or need for including the process for early review of site suitability issues in part 53.

Much of the subpart H regulatory text is identical to the corresponding language in parts 50 and 52, with minor changes to account for cross-references in part 53, to make language technology neutral, or to reflect the unique analytical approach in part 53. In these instances, this preamble discussion will describe the language as “equivalent” to the existing corresponding requirement in part 50 or part 52 and will describe any deviations, where applicable.

Because part 53 carries over the majority of the licensing options from parts 50 and 52, there are several sections in subpart H that are similar to existing regulations in parts 50 and 52. Section 53.1100 addresses filing of applications for licenses, certifications, or approvals under oath or affirmation and is equivalent to Sec. 50.30. Section 53.1100 does not include the current requirement in Sec. 50.30(a)(2) that the applicant maintain the capability to generate additional copies, because it is unnecessary in the age of electronic submissions. In addition, the existing requirement on applications for OLs in Sec. 50.30(d) is included in Sec. 53.1124(g)(2), “Relationship between sections,” covering OLs, rather than in Sec. 53.1100. Section 53.1100(a)(1) also includes filing requirements equivalent to those in Sec. Sec. 52.15, 52.45, 52.135, and 52.151. Section 53.1101 lays out activities requiring an NRC license and is equivalent to Sec. 50.10(b). Section 53.1103 addresses combining applications and is equivalent to Sec. Sec. 50.31, 50.52, and 52.8. Section 53.1103(b) continues the Commission's practice of combining multiple authorizations for a facility under parts 30, 40, 50, 52, and 70 into one license based on the Commission's authority under section 161h of the AEA to combine NRC licenses. Section 53.1106 addresses elimination of repetition and is equivalent to Sec. 50.32.

Section 53.1109 provides general information requirements for the content of applications submitted to the NRC under part 53 and is equivalent to Sec. 50.33, with the exception of Sec. 50.33(f) on financial qualifications, which is covered in subpart J, and Sec. 50.33(h) on earliest and latest dates for completion of construction, which is covered in Sec. 53.1306 of subpart H. Each application must include information to address the items in Sec. 53.1109 as cited in the appropriate section of subpart H for the application type.

One change from current requirements can be found in Sec. 53.1109(i), which is not limited to electricity generation as it is currently in part 50. Some prospective NRC applicants are considering development of nuclear plants for other commercial ventures, such as process heat generation or hydrogen production. In addition, Sec. 53.1109(j), which requires applications containing classified information to separate that information from the unclassified information in the application, refers to “Restricted Data or classified National Security Information” instead of the term used in the corresponding provision in Sec. 50.33(j), “Restricted Data or other defense information.” This change was made to use the defined term in 10 CFR part 95, “Facility Security Clearance and Safeguarding of National Security Information and Restricted Data,” rather than “defense information” as used in Sec. 50.33(j). The usage in Sec. 50.33(j) dates back to the Atomic Energy Commission amendment of that section on January 19, 1956 (21 FR 355, 357), and was not changed with the issuance of part 95 (45 FR 14476; March 5, 1980) after the establishment of the NRC and the 1975 reissuance of the former Atomic Energy Commission regulations. The revised terminology also aligns with its usage in Sec. 53.1115.

Section 53.1112 addresses environmental conditions and is equivalent to Sec. 50.36b. Section 53.1115 addresses requirements for agreements limiting access to classified information and is equivalent to Sec. 50.37.

Section 53.1118 addresses ineligibility of certain applicants and is similar to Sec. 50.38 but has been revised consistent with section 301 of the ADVANCE Act. That section of the ADVANCE Act designates certain exceptions from the foreign ownership, control, or domination (FOCD) provision set forth in the AEA. Specifically, section 301 states that if the Commission determines that issuance of the applicable license to that entity is not inimical to the common defense and security or public health and safety, then the FOCD restriction shall not apply to an entity that is owned, controlled, or dominated by: (1) the government of a country that is a member of the Organisation for Economic Co-operation and Development or the Republic of India; (2) a corporation that is incorporated in one of those countries; or (3) a citizen or national of one of those countries, subject to some additional exclusions. Those additional exclusions are based on whether any members of the excepted countries were on certain sanctions lists on the ADVANCE Act's date of enactment.

Section 53.1120 addresses exceptions and exemptions from licensing requirements for Department of Defense and DOE facilities and is equivalent to Sec. 50.11. Section 53.1121 addresses public inspection of applications and is equivalent to Sec. 50.39.

Section 53.1124 addresses the relationship between the various licenses, certifications, and approvals provided in this subpart, and the requirements are equivalent to a number of similar provisions in parts 50 and 52, including Sec. Sec. 50.10, 52.13, 52.43, 52.73, 52.133, and 52.153. New provisions are provided in Sec. 53.1124(c) and (d) that allow an application for either a

standard design approval or a standard DC under part 53 to reference applicable licensing-basis information that supported issuance of an OL or COL under part 53. These provisions will offer additional flexibility beyond what is currently allowed under parts 50 or 52 for an applicant who may wish to license a first-of-a-kind reactor for operation prior to seeking generic approval or certification of the standard design.

Section 53.1124(e) addresses the limitation that a manufactured reactor may only be transported domestically to a site with a COL or CP and is generally equivalent to Sec. 52.153. The NRC has not included specific requirements within part 53 directing how a CP application referencing an ML must be structured, including how the ITAAC required for an ML under Sec. 53.1282 is to be addressed by a CP applicant. Instead, part 53 leaves the matter open to possible approaches to be addressed in future regulatory guidance or proposed by future applicants. Section 53.1124(e) includes an additional statement to make it clear that a manufactured reactor may be exported in accordance with part 110.

Section 53.1130 addresses LWAs and is equivalent to Sec. 50.10.

Sections 53.1140 through 53.1188 address applications for, issuance of, and other provisions related to ESPs under part 53. Section 53.1140 describes how the contents of Sec. Sec. 53.1140 through 53.1188 address ESPs and is equivalent to Sec. 52.12. Section 53.1144 addresses general information requirements for the content of applications and is equivalent to Sec. 52.16.

Section 53.1146 specifies requirements for the technical contents of applications and is equivalent to Sec. 52.17. Section 53.1146(b)(2) provides applicants for ESPs a regulatory option to propose major features of the emergency plans or complete integrated emergency plans in accordance with either the requirements in Sec. 50.160 of this chapter, or the requirements in appendix E to part 50 of this chapter and Sec. 50.47(b) of this chapter, as applicable.

Section 53.1149 addresses standards for review of ESP applications and administrative review of applications, including hearings, and is equivalent to Sec. Sec. 52.18 and 52.21. Section 53.1155 addresses referral to the ACRS and is equivalent to Sec. 52.23. Section 53.1158 addresses issuance of ESPs and is equivalent to Sec. 52.24. Section 53.1161 addresses the extent of activities permitted and is equivalent to Sec. 52.25. Section 53.1164 addresses the duration of an ESP and is equivalent to Sec. 52.26. Section 53.1167 addresses provisions for requesting an LWA after issuance of an ESP and is equivalent to Sec. 52.27. Section 53.1170 addresses transfers of ESPs and is equivalent to Sec. 52.28. Section 53.1173 addresses applications for ESP renewals and is equivalent to Sec. 52.29, although the final rule removes the requirement to refer the renewal to the ACRS consistent with current agency practice. Section 53.1176 addresses criteria for renewal of an ESP and is equivalent to Sec. 52.31. Section 53.1179 addresses the duration of an ESP renewal and is equivalent to Sec. 52.33. Section 53.1182 addresses the use of a site for purposes other than those described in the permit and is equivalent to Sec. 52.35. Section 53.1188 addresses finality of ESP determinations and is equivalent to Sec. 52.39.

Sections 53.1200 through 53.1221 address applications for, issuance of, and other provisions related to standard design approvals under part 53. Section 53.1200 describes how the contents of Sec. Sec. 53.1200 through 53.1221 address standard design approvals and is equivalent to Sec. 52.131. Section 53.1206 addresses general information requirements for the content of applications and is equivalent to Sec. 52.136.

Section 53.1209 addresses requirements for the technical content of applications and is largely equivalent to Sec. 52.137. In Sec. 53.1209(a), the NRC includes text that expands the discussion of a “major portion” of standard design approvals. Additional discussion regarding standard design approvals for a major portion of a standard design can be found in the NRC's “A Regulatory Review Roadmap for Non- Light Water Reactors,” which considers the Nuclear Innovation Alliance report “Clarifying `Major Portions' of a Reactor Design in Support of a Standard Design Approval.” Section 53.1209(b) outlines the required content of the Final Safety Analysis Report (FSAR). Requirements in Sec. 53.1209(b)(2) for portions of the application addressing design information state that the application must include design information equivalent to that required for a standard DC. This reference to the pertinent DC requirements (specifically, those in Sec. 53.1239(a)(2) through (27)) is an efficiency that prevents the need to repeat many of the same requirements for the content of a standard design approval application.

Section 53.1210 addresses requirements for the content of a standard design approval application other than the FSAR. Section 53.1210(a) requires the inclusion of a description of availability controls that are not included in the FSAR.

Section 53.1212 addresses standards for review of applications and is equivalent to Sec. 52.139. Section 53.1215 addresses referral to the ACRS and is equivalent to Sec. 52.141. Section 53.1218 addresses staff approval of designs and duration of design approvals and is equivalent to Sec. Sec. 52.143 and 52.147. Section 53.1221 addresses finality of standard design approvals and information requests and is equivalent to Sec. 52.145 with the exception that it extends such finality to a standard approval referenced in a DC application. Standard design approvals issued to date under part 52 have been issued during the NRC's review of the standard DC application and have relied on the same application content. However, a future scenario could arise where the DC application is not submitted until after a design approval has been granted. The NRC would apply the same finality provisions in this situation as in the situation where a standard design approval is referenced in a COL application.

There is no equivalent to Sec. 53.1221(d) in part 52 for standard design approvals. This provision states that the Commission will require, before granting a CP, COL, OL, or ML that references a standard design approval, that information normally contained in engineering documents be completed and available for audit. A similar provision is included in part 52 in relation to a standard DC; and the NRC would require that design and analysis information needed for the Commission to make its safety determination be complete and available for any application the NRC is reviewing. Making this explicit provides increased clarity to future standard design approval applicants under part 53.

Sections 53.1230 through 53.1263 address applications for, issuance of, and other provisions related to standard DCs under part 53. Section 53.1230 addresses general provisions for standard DCs and is equivalent to Sec. 52.41. Section 53.1236 addresses general information requirements for the content of applications and is equivalent to Sec. 52.46. Section 53.1239 addresses requirements for the technical content of applications and is equivalent to Sec. 52.47(a). The requirements in Sec. 53.1239 have been modified from the analogous requirements in Sec. 52.47(a) to align with the technical requirements in part 53.

Section 53.1241 addresses requirements for the content of a standard DC application other than the

FSAR and is equivalent to Sec. 52.47(b) and (d).

Section 53.1242 addresses review of applications and is equivalent to Sec. Sec. 52.48 and 52.51. Section 53.1242(c) includes a provision that allows a DC applicant to reference applicable licensing-basis information for an OL or COL issued under part 53. As explained previously, this provision explicitly allows flexibility for an applicant who may wish to license a first-of-a-kind reactor for operation prior to seeking certification of the generic reactor design. For NRC findings on a reactor design in an OL or COL proceeding, this provision provides finality in a subsequent DC application that references information on the OL or COL proceeding's docket. This finality accorded to the OL or COL findings would bind the NRC staff and the ACRS but would not bind members of the public or the Commission. (To the extent an Atomic Safety and Licensing Board (ASLB) might have a role in a DC rulemaking, the OL or COL findings would not bind the ASLB either.) Specifically, members of the public would have the opportunity to comment on a proposed DC rule under well-established NRC practice. The rationale for binding the NRC staff and ACRS is similar to the rationale for a COL applicant referencing a standard design approval under part 52.

Section 53.1245 addresses referral to the ACRS and is equivalent to Sec. 52.53. Section 53.1248 addresses issuance of standard DCs and is equivalent to Sec. 52.54. Section 53.1251 addresses duration of certifications and is equivalent to Sec. 52.55. Section 53.1254 addresses application for renewal and is equivalent to Sec. 52.57, although the final rule removes the requirement to refer the renewal to the ACRS consistent with current agency practice. Section 53.1257 addresses criteria for renewal and is equivalent to Sec. 52.59. Section 53.1260 addresses duration of renewals and is equivalent to Sec. 52.61. Section 53.1263 addresses finality of standard DCs and is equivalent to Sec. 52.63.

Sections 53.1270 through 53.1291 address applications for, issuance of, and other provisions related to MLs covering manufacturing activities at one or more licensee facilities under part 53. Section 53.1270 addresses the scope of these sections and is equivalent to Sec. 52.151.

Section 53.1276 addresses general information requirements for the content of ML applications and is equivalent to Sec. 52.156, with one exception. Section 53.1276 requires each application for an ML to also include the information required by Sec. 53.1109(e). This information includes the type of license applied for, the use to which the facility will be put, the period of time for which the license is sought, and a list of other licenses, except operator's licenses, issued or applied for in connection with the proposed facility to address the potential variations in how MLs might be formulated under part 53.

Section 53.1279 addresses requirements for the technical content of applications for MLs to be included in the FSAR and is equivalent to Sec. 52.157. In addition, the requirements in Sec. 53.1279(a) and (b) have been modified from the analogous requirements in Sec. 52.157 to align with the technical requirements in part 53. Section 53.1279(a)(2) outlines the required content of the application addressing design information and states that the application must include design information equivalent to that required for a standard DC. This reference to the pertinent DC requirements is an efficiency that prevents the need to repeat the same requirements for the content of an ML application.

Section 53.1279(c) provides application requirements related to the deployment of the completed manufactured reactor. Section 53.1279(c)(1) requires inclusion of information related to the procedures governing the preparation of the manufactured reactor for shipping to the site where it is to be operated, the conduct of shipping, and the verification of the condition of the shipped items upon receipt at the site. Section 53.1279(c)(2) requires that the application include information on the interaction of the design, manufacture, and installation of a manufactured reactor within the applicant's organization and the manner by which the applicant will ensure close integration between the designer, contractors, and any licensee of a facility in which the manufactured reactor is to be installed. Finally, Sec. 53.1279(c)(3) requires that the application include a description of the measures used for the control of interfaces between the holder of the ML and the holder of the COL or CP for the commercial nuclear plant at which the manufactured reactor is to be installed. This information is necessary for the NRC to determine whether the applicant has appropriate controls in place to ensure coordination between parties involved in the design, manufacture, and eventual operation of any reactor manufactured under an ML.

Section 53.1279(d) includes additional requirements for application content for applicants seeking an ML for manufactured reactors that will be fueled at the factory under a part 70 license, consistent with the requirements in Sec. 53.620(d). These provisions require the application to include information related to loading fuel and the required features to prevent criticality and to otherwise provide assurance that the fueled manufactured reactor can be successfully transported, installed, and operated at a site for which the Commission has issued a COL or CP and OL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor.

Section 53.1282 provides requirements for other application content for MLs and is equivalent to Sec. 52.158. Section 53.1282(a)(1) provides requirements to include in the ML application the ITAAC within the scope of the ML that the COL or CP holder referencing the ML must satisfy. Section 53.1282(a)(2) requires that the ITAAC from a referenced standard design apply to the portions of the ML design within the scope of the referenced standard design. Section 53.1282(a)(3) states that a COL application may include a notification that required referenced standard DC ITAAC have been satisfied at the manufacturing facility.

Section 53.1282(b) requires an ML application to include an environmental report and, consistent with existing requirements, Sec. 53.1282(b)(2) notes that if the ML application references a standard DC, the environmental report need not contain a discussion of severe accident mitigation design alternatives for the manufactured reactor as used in a commercial nuclear plant.

Section 53.1285 provides standards for review of applications and administrative review of applications for MLs, including hearings, and is equivalent to Sec. Sec. 52.159 and 52.163.

Section 53.1286 addresses referral of applications to the ACRS and is equivalent to Sec. 52.165. Section 53.1287 addresses issuance of an ML and is equivalent to Sec. 52.167.

Section 53.1288 addresses finality of MLs and is equivalent to Sec. 52.171, except that part 53 does not include the constraint that the Commission may only grant a request for a departure from an ML for an applicant who references or uses a manufactured reactor if special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. This is consistent with the differences in the allowance for changes to a manufactured reactor in part 53 (as noted in the discussion of Sec. Sec. 53.1530 and 53.1550 in this

document) as compared to part 52. Section 53.1291 provides for a 40- year duration for an ML, consistent with the duration provided for a DC under Sec. 53.1251. Section 53.1293 addresses the transfer of MLs and is equivalent to Sec. 52.175. Section 53.1295 addresses the renewal of MLs and is equivalent to Sec. Sec. 52.177, 52.179 and 52.181, with minor exceptions.

Section 53.1295(a)(3) states that an ML for which a timely application for renewal has been filed remains in effect until the Commission has made a final determination on the renewal application. However, this provision omits a limitation from the equivalent provision in Sec. 52.177 which prohibits the holder of an ML from beginning the manufacture of a manufactured reactor less than 3 years before the expiration of the license. This limitation was omitted in part 53 because future reactor applicants may present smaller, simpler designs, to include microreactor designs, in ML applications than those that were envisioned when the existing requirements were written. Eliminating the 3-year constraint in this provision will provide greater flexibility for ML holders related to manufactured reactors being produced close to the time when the ML expires. Additionally, Sec. 53.1295(c) provides for a 40-year term for a renewed ML, consistent with the term for an initial ML under Sec. 53.1291. Finally, the final rule removes the requirement to refer the renewal to the ACRS consistent with current agency practice.

Sections 53.1300 through 53.1348 address applications for, issuance of, and other provisions related to CPs under part 53. Section 53.1300 sets out general requirements for CPs and is equivalent to Sec. 50.23. Section 53.1306 addresses the general information requirements for the content of applications for CPs and is similar to Sec. 50.33(f) and (h). However, the requirements for demonstrating financial qualification are different for part 53 applicants than the existing requirements for applicants under part 50 or 52. The part 53 requirements do not include the existing requirements under part 50 or 52 for an applicant to provide information to demonstrate that it “possesses or has reasonable assurance of obtaining” the funds necessary for construction and operation along with associated financing details. Instead, part 53 replaces that requirement with a requirement to provide information that demonstrates that the applicant “appears to be financially qualified,” similar to the standard used in Sec. 70.23(a)(5).

Section 53.1309 addresses requirements for the technical content of applications for CPs and includes the requirement to submit a Preliminary Safety Analysis Report (PSAR) that describes the facility and presents a preliminary safety analysis of the facility as a whole. This is in contrast to an OL application, which is required to include an FSAR that describes the facility and presents a final safety analysis of the facility as a whole. Section 53.1309 is equivalent to Sec. 52.17(a)(1)(iv) through (a)(1)(x) and 52.17(b), with two exceptions. First, Sec. 53.1309 replaces the analysis of the dose criteria required by Sec. 52.17(a)(1)(ix) with analysis to demonstrate compliance with the safety criteria defined in Sec. Sec. 53.210 and 53.220. Second, Sec. 53.1309(a)(2) adds a requirement for a CP application to include several categories of detailed design information, although Sec. 53.1309(a)(2)(ii) allows certain relaxations of this requirement in view of aspects of a design that may not yet be fully developed. Section 53.1309 references the requirements for the content of an ESP application to address application requirements related to siting and references the requirements for the content of a DC application to address application requirements related to design of the commercial nuclear plant. Section 53.1309(a)(2)(ii) addresses the treatment of preliminary design information and notes that information provided in the application may include some aspects of the design that are not fully developed. This provision requires that the completed design, including any changes during construction, be described in the FSAR in an application for an OL. This includes the requirement for a description of the PRA, other SREs, or a combination thereof required by Sec. 53.450(a) and its results. Probabilistic risk assessments, other SREs, or a combination thereof developed for commercial nuclear plants prior to construction are based on the design and other information available at the time of the CP application. PRAs performed in early design stages or prior to construction may be inherently less detailed and may include projected information that will be subsequently verified or revised when the plant is built. Section 53.1309(a)(4) addresses preliminary description of the plans for coping with emergencies.

Section 53.1312 addresses other application content for CPs. Section 53.1312(a)(1) is equivalent to Sec. 52.80(b) but is adapted for a CP application. Section 53.1312(a)(2) is equivalent to Sec. 52.80(c) but is adapted for a CP application. Section 53.1312(b)(1) is equivalent to Sec. 52.79(b), (c), and (d) but is adapted for a CP application. Section 53.1312(b)(2) is equivalent to portions of Sec. Sec. 52.63(b)(1), 52.79(b)(1) through (b)(3), (c), and (d)(1) and (d)(3), 52.80, and 52.93(b) but is adapted for a CP application. Guidance for equivalent requirements in parts 50 and 52 is also addressed in RG 1.206, “Applications for Nuclear Power Plants,” Revision 1, section C.1.7.

Section 53.1315 addresses standards for review of applications and administrative review of applications, including hearings, and is equivalent to Sec. Sec. 52.81 and 52.85 but is adapted for a CP application.

Section 53.1318 addresses finality of NRC approvals, licenses, and certifications referenced in a CP application and is equivalent to Sec. 52.83(a) but is adapted for a CP application.

Section 53.1324 addresses referral to the ACRS and is equivalent to Sec. 50.58(a) and to Sec. 52.87 but is adapted for a CP application.

Section 53.1327 addresses authorization to conduct LWA activities and is equivalent to Sec. 52.91 but is adapted for a CP application. Section 53.1327(a) is equivalent to Sec. 52.91(a) but is adapted for a CP application. Section 53.1327(b) is equivalent to Sec. 52.91(b) but is adapted for a CP application. Section 53.1330 addresses exemptions, departures, and variances for CP applicants.

Section 53.1333 addresses issuance of CPs. Section 53.1333(a) is equivalent to Sec. 50.35(a). Section 53.1333(b) is equivalent to Sec. 50.35(b) and to Sec. 52.97(c) but is adapted for a CP application. Section 53.1336 addresses the effect of CPs and is equivalent to Sec. 50.35(b). Section 53.1342 addresses the duration of CPs. Section 53.1342(a) is equivalent to Sec. 50.55(a). Section 53.1342(b) is equivalent to Sec. 50.55(b). Section 53.1345 addresses the transfer, assignment, and disposal of CPs and is equivalent to Sec. 50.80. Section 53.1348 addresses the termination of CPs and is equivalent to Sec. Sec. 52.3(b)(8) and 52.110(a)(1) but is adapted for a CP application.

Sections 53.1360 through 53.1405 address applications for, issuance of and other provisions related to OLs under part 53.

Section 53.1366 addresses requirements for the general content of applications for OLs. It refers to general content requirements in Sec. 53.1109 and requires supplemental information. Section 53.1366 is similar to Sec. 50.33(f). However, the requirements for demonstrating financial qualification

are different for part 53 applicants than the existing requirements for applicants under part 50 or 52. The part 53 requirements do not include the existing requirements for an applicant to provide information to demonstrate that it “possesses or has reasonable assurance of obtaining” the funds necessary for construction and operation along with associated financing details. Instead, part 53 replaces that requirement with a requirement to provide information that demonstrates that the applicant “appears to be financially qualified,” similar to the standard used in Sec. 70.23(a)(5).

Section 53.1369 provides requirements for the technical content of applications for OLs to be included in the FSAR and is equivalent to Sec. 50.34(b) but has been modified to align with the technical requirements in part 53. It requires that the FSAR include and, as needed, update information provided in the PSAR that was submitted and reviewed to support the associated CP application.

Similar to the requirements for the content of CP applications, Sec. 53.1369(a) references the requirements for the content of an ESP application to address application requirements related to the site. Section 53.1369(b) references the requirements for the content of a DC application to address some of the application requirements related to design of the commercial nuclear plant.

Section 53.1369(d) requires a description of the Integrity Assessment Program that is required by Sec. 53.870. Section 53.1369(e) is equivalent to Sec. 50.34(e). Section 53.1369(g) provides requirements for OL application content to support Sec. 53.730 related to the role of personnel in the operation of the commercial nuclear plant and is adapted from requirements in part 55 and Sec. 50.34(f). Likewise, Sec. 53.1369(h) provides requirements for OL application content related to training programs to support Sec. Sec. 53.730(g) and 53.830 and includes requirements equivalent to Sec. 50.34(b)(8), Sec. 52.79(a)(33), and part 55. Section 53.1369(i) provides requirements for OL application content related to emergency plans to support Sec. 53.855 and is equivalent to Sec. 50.34(b)(6)(v).

Section 53.1369(j) provides requirements for OL application content related to the applicant's organizational structure and is equivalent to Sec. 50.34(b)(6)(i). Section 53.1369(k) provides requirements for OL application content related to the applicant's proposed maintenance program to support Sec. 53.715 and is equivalent to Sec. 50.34(b)(6)(iv). Section 53.1369(l) provides requirements for OL application content related to the applicant's quality assurance program to support Sec. 53.865 and is equivalent to Sec. 50.34(b)(6)(ii). Section 53.1369(m) provides requirements for OL application content related to the applicant's proposed radiation protection program to support Sec. 53.850 and is equivalent to Sec. 50.34(b)(3).

Sections 53.1369(n) through (p) provide requirements for OL application content related to the applicant's proposed physical security program to support Sec. 53.860(a) and are equivalent to Sec. 50.34(c) and (d). Section 53.1369(q) provides requirements for OL application content related to the applicant's proposed cybersecurity plan to support Sec. 53.860(d) and is equivalent to Sec. Sec. 52.79(a)(36)(iv) and 73.54. Section 53.1369(r) provides requirements for OL application content related to the implementation of proposed security, safeguards, and cybersecurity plans to support Sec. 53.860 and is equivalent to Sec. 52.79(a)(35)(ii) and 52.79(a)(36)(iv) and (v).

Section 53.1369(s) provides requirements for OL application content related to the applicant's proposed fire protection program to support Sec. 53.875 and is equivalent to Sec. 52.79(a)(40). Section 53.1369(t) provides requirements for OL application content related to the applicant's proposed ISI and IST program to support Sec. 53.880 and is equivalent to part of Sec. 52.79(a)(11). Section 53.1369(w) provides requirements for OL application content related to the applicant's general employee training program to support Sec. 53.830 and is equivalent to Sec. 52.79(a)(33). Section 53.1369(x) provides requirements for OL application content related to the applicant's FFD program to support part 26 and is equivalent to Sec. 52.79(a)(44). Section 53.1369(y) provides requirements for OL applicants' programs to demonstrate that any safety questions identified at the CP stage have been resolved and is equivalent to Sec. 50.34(b)(5). Section 53.1369(z) provides requirements for OL applicants to describe how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof to support Sec. 53.440(a). It is largely equivalent to Sec. Sec. 50.34(b)(5) and 50.43(e). Section 53.1369(aa) provides requirements for OL application content related to the applicant's proposed TS to support Sec. 53.710(a) and is equivalent to Sec. 50.34(b)(6)(vi).

Section 53.1372 addresses requirements for the content of OL applications other than the FSAR. Section 53.1372(a) requires submission of an environmental report and is equivalent to Sec. 50.30(f) and Sec. 51.53(b). Section 53.1372(b) does not have a direct parallel in parts 50 and 52 and requires the inclusion of a description of availability controls that are not included in the FSAR to support Sec. 53.710(b).

Section 53.1375 addresses standards for review of OL applications and the administrative review of applications, including hearings, and is equivalent to Sec. Sec. 52.81 and 52.85, except that the NRC has omitted 10 CFR part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” from the list of standards in Sec. 53.1375(a). Part 53 does not include detailed requirements related to renewal of licenses, although a general provision and possible placeholder for future requirements has been included as Sec. 53.1595. The NRC will decide after the part 53 final rule is published whether this section will be retained in part 53 to address license renewal or whether the agency will take another approach to address license renewal for part 53 licensees, such as amending part 54 to address part 53 licensees.

Section 53.1381 addresses referral to the ACRS and is equivalent to Sec. Sec. 50.58 and 52.87. Section 53.1384 addresses exemptions, departures, and variances for OL applicants. Section 53.1384(a) is equivalent to Sec. 52.93 but is adapted for OLs. Section 53.1384(b) is equivalent to Sec. Sec. 52.39(d) (with respect to ESPs) and 52.93 but is adapted for OLs.

Section 53.1387 addresses issuance of OLs. The introductory paragraph is equivalent to Sec. 50.56. Section 53.1387(a)(1)(i) is equivalent to Sec. Sec. 50.50 and 50.57(a)(1). Section 53.1387(a)(1)(ii) is equivalent to Sec. 50.50. Section 53.1387(a)(1)(iii) is equivalent to Sec. 50.57(a)(2). Section 53.1387(a)(1)(iv) is equivalent to Sec. 50.57(a)(3). Section 53.1387(a)(1)(v) is equivalent to Sec. 50.57(a)(4). Section 53.1387(a)(1)(vi) is equivalent to Sec. 50.57(a)(6). Section 53.1387(a)(1)(vii) is equivalent to Sec. 50.57(a)(5). Section 53.1387(a)(1)(viii) is equivalent to Sec. 52.97(a)(1)(vi) but is adapted for OLs. Section 53.1387(c) is equivalent to Sec. 50.57(b). Section 53.1387(d) is equivalent to Sec. Sec. 50.36(b) and 50.50.

Section 53.1390 addresses backfitting of OLs and is equivalent to Sec. 52.98(a) but adapted for an OL application. Section 53.1396 addresses duration of an OL and is equivalent to Sec. 50.51(a) and Sec. 52.104. Section 53.1399 addresses transfer, assignment, and other

disposition of an OL and is equivalent to Sec. 50.80. Section 53.1402 addresses applications for renewal of an OL and refers to Sec. 53.1595. Section 53.1405 addresses continuation of an OL and is equivalent to Sec. 52.109 but is adapted to address an OL.

Sections 53.1410 through 53.1461 address applications for, issuance of, and other provisions related to COLs under part 53. Section 53.1410 describes the contents of Sec. Sec. 53.1410 through 53.1461 and is equivalent to Sec. 52.71. Section 53.1413 addresses general information requirements for the content of applications for COLs and is equivalent to Sec. 52.77, which references Sec. 50.33. Most of the provisions from Sec. 50.33 applicable to COLs are restated in Sec. 53.1109. However, the requirements for demonstrating financial qualification are different for part 53 applicants than the existing requirements for applicants under part 50 or 52. The part 53 requirements do not include the existing requirements under part 50 or 52 for an applicant to provide information to demonstrate that it “possesses or has reasonable assurance of obtaining” the funds necessary for construction and operation along with associated financing details. Instead, part 53 replaces that requirement with a requirement to provide information that demonstrates that the applicant “appears to be financially qualified,” similar to the standard used in Sec. 70.23(a)(5).

Section 53.1416 addresses the technical content to be included in an FSAR for an application for a COL and is equivalent to Sec. 52.79 except as modified to reflect the technical requirements in part 53 and with one addition. Section 53.1416 includes the statement that the Commission will require, before issuance of a COL, that information normally contained in engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination. This statement is equivalent to DC application requirements in Sec. 52.47 and is included in Sec. 53.1416 for clarity.

Similar to the requirements for the content of OL applications, Sec. 53.1416(a)(1) references the requirements for the content of an ESP application to address application requirements related to siting. Section 53.1416(a)(2) references the requirements for the content of a DC application to address some of the application requirements related to design of the commercial nuclear plant. The remaining items under Sec. 53.1416(a) are likewise similar to the required content for OL applications under Sec. 53.1369(a). Section 53.1416(b) requires COL applicants to provide a report documenting the resolution of any safety questions for SSCs for which research and development was necessary to confirm the adequacy of their design and is equivalent to Sec. 50.34(b)(5). Section 53.1416(c) provides requirements for COL applicants to describe how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof to support Sec. 53.440(a). It is largely equivalent to Sec. Sec. 52.79(a)(24) and 50.43(e). Section 53.1416(d) addresses the content of COL applications referencing an ESP. Section 53.1416(e) addresses the content of COL applications referencing a standard design approval. Section 53.1416(f) addresses the content of COL applications referencing a standard DC. Section 53.1416(g) addresses the content of COL applications referencing an ML.

Section 53.1419 addresses other application content for COLs and is equivalent to Sec. 52.80. Section 53.1419(a)(2) is new and requires the inclusion of a description of availability controls that are not required to be included in the FSAR.

Section 53.1422 addresses standards for review of applications and the administrative review of applications, including hearings, and is equivalent to Sec. Sec. 52.81 and 52.85. The NRC has removed part 54 from the list of standards in Sec. 53.1422(a). Part 53 does not include requirements related to renewal of licenses, in relation to Sec. Sec. 53.1422 and 53.1595.

Section 53.1425 addresses the finality of NRC approvals referenced in a COL application and is equivalent to Sec. 52.83(a). Section 53.1431 addresses the referral of COL applications to the ACRS for review and is equivalent to Sec. 52.87. Section 53.1434 addresses the authorization to conduct LWA activities and is equivalent to Sec. 52.91. Section 53.1437 addresses exemptions, departures, and variances and is equivalent to Sec. 52.93. Section 53.1440 addresses issuance of COLs and is equivalent to Sec. 52.97. Section 53.1443 addresses finality of COLs and is equivalent to Sec. 52.98.

Section 53.1449 addresses inspection during construction and is equivalent to Sec. 52.99. Section 53.1452 addresses operation under a COL and is equivalent to Sec. 52.103. Paragraph Sec. 53.1452(a) includes footnotes to provide that, for licensees installing fueled manufactured reactors under a COL, (1) the COL holder will notify the NRC of its scheduled date for initiating the removal of the features to prevent criticality required under Sec. 53.620(d)(1) rather than its scheduled date for the initial loading of fuel, and (2) the NRC will time its publication of the notice of intended operation based on the COL holder's schedule for initiating the removal of the features to prevent criticality required under Sec. 53.620(d)(1) rather than the COL holder's scheduled date for the initial loading of fuel. These footnotes are consistent with the provisions of Sec. 53.620(d)(1)(iv), which state that, upon initiating the removal of the features to prevent criticality in the manufactured reactor's place of operation, the fueled manufactured reactor has commenced operation. For reactors without the features to prevent criticality under Sec. 53.620(d)(1), operation begins with initial fuel load. In both cases, removal of the physical features to prevent criticality (for reactors with such features) and initial fuel load (for reactors without such features) put a fully constructed utilization facility in a position to sustain a nuclear chain reaction, and in both cases, the utilization facility cannot sustain a nuclear chain reaction (for lack of sufficient reactivity) until the action takes place. Therefore, the NRC believes that initiating the removal of the features to prevent criticality is the best analogue to initial loading of fuel for reactors without such features.

The footnote in Sec. 53.1452(a) regarding timing of the notice of intended operation for fueled manufactured reactors with features to prevent criticality also addresses the requirements of section 189a.(1)(B)(i) of the AEA. This section requires, in part, that “[n]ot less than 180 days before the date scheduled for initial loading of fuel into a plant by a licensee that has been issued a combined construction permit and operating license under section 185b., the Commission shall publish in the Federal Register notice of intended operation.” That section further requires that this notice provide a 60-day period in which to request a hearing “on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria of the license.” In the case where a fueled manufactured reactor arrives at the site where it is to be operated by a COL holder, the manufacturer would have

loaded fuel at the factory under its part 70 license. Therefore, at the site of operation, there would not be “initial loading of fuel into a plant by a licensee that has been issued a combined construction permit and operating license” (emphasis added). Under a literal reading of the entry condition in Act section 189a.(1)(B)(i), this situation would not trigger its requirements. However, the purpose of the provision is to offer the hearing opportunity at least 180 days prior to when the fuel is loaded and ready for use at its authorized location. It would be contrary to that purpose if, in this situation, the Commission did not publish the notice of intended operation and opportunity for the public to request a hearing on conformance with the acceptance criteria in the COL for the site of operation. To fulfill the underlying purpose of the law, the NRC is timing the notice of intended operation based on the COL holder's schedule for initiating the removal of the features to prevent criticality required under Sec. 53.620(d)(1). This action by the COL holder is the best analogue to initial fuel load by the COL holder for the reasons stated previously. This analogue is adopted in other sections of part 53 and related sections in parts 50 and 73 that use initial fuel loading to identify a transition point for the applicability of regulatory requirements. To address the possible loading of fuel into a manufactured reactor for subsequent transport to and use at a commercial nuclear plant, multiple sections that determine the applicability of regulations have been developed or revised to allow for either initial fuel load or initiating the removal of the features to prevent criticality required under Sec. 53.620(d)(1) for a fueled manufactured reactor to determine the applicability of the requirement, as appropriate.

Section 53.1455 addresses duration of COL and is equivalent to Sec. 52.104. Section 53.1456 addresses the transfer of a COL and is equivalent to Sec. 52.105. Section 53.1458 addresses application for renewal and is equivalent to Sec. 52.107. Section 53.1461 addresses continuation of COL and is equivalent to Sec. 52.109.

Section 53.1470 addresses standardization of commercial nuclear plant designs and licenses to construct and operate commercial power reactors of identical design at multiple sites and is equivalent to appendix N to part 52, with one exception. Paragraph Sec. 53.1470(b) provides flexibility regarding the timing of the applications to be treated together under Sec. 53.1470. Each application can either list the other applications or specify that such other applications will be submitted to the NRC within 12 months of submittal of the first application. This section sets out the particular requirements and provisions applicable to situations in which applications for CPs and subsequent OLs, or COLs, under this part are filed by one or more applicants for licenses to construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple sites. Additional information related to this section is provided in the final rule to revise part 52 (72 FR 49352; August 28, 2007).