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

Sec. 53.400 Design features for licensing-basis events.

(a) Design features must be provided for each commercial nuclear plant such that, when combined with corresponding human actions and programmatic controls, the plant will satisfy the safety criteria defined in Sec. Sec. 53.210 and 53.220.

(b) Design features must ensure that the safety functions identified in Sec. 53.230 are fulfilled during licensing-basis events (LBEs).

Sec. 53.410 Functional design criteria for design-basis accidents.

(a) Functional design criteria must be defined for each design feature classified as safety-related (SR) in terms of its role in demonstrating compliance with the safety criteria defined in Sec. 53.210.

(b) The identification of special treatments associated with the design of SR structures, systems, and components (SSCs) must consider human actions and programmatic controls identified and implemented in accordance with this and other subparts to achieve and maintain the reliability and capability of SSCs relied upon to satisfy the defined functional design criteria and the safety criteria required in Sec. 53.210, and to maintain consistency with analyses required by Sec. 53.450(f).

Sec. 53.415 Protection against external hazards.

Safety-related SSCs must be protected against or must be designed to withstand the effects of natural phenomena (e.g., earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and constructed hazards (e.g., dams, transportation routes, military and industrial facilities) considering an event severity up to the design-basis external hazard levels as determined under Sec. 53.510 without losing the capability to perform the safety functions identified under Sec. 53.230. Specific requirements for earthquake engineering are included in Sec. 53.480.

Sec. 53.420 Functional design criteria for licensing-basis events other than design-basis accidents.

(a) Functional design criteria must be defined for each design feature classified as SR or non-safety-related but safety-significant (NSRSS) in terms of its role in demonstrating compliance with--

(1) The safety criteria in Sec. 53.220; and

(2) The evaluation criteria in Sec. 53.450(e).

(b) The identification of special treatments associated with the design of SR and NSRSS SSCs must consider human actions and programmatic controls identified and implemented in accordance with this and other subparts to achieve and maintain the reliability and capability of SSCs relied upon to satisfy--

(1) The safety criteria in Sec. 53.220; and

(2) The evaluation criteria in Sec. 53.450(e).

Sec. 53.425 Design features and functional design criteria for normal operations.

(a) Design features must be provided for each commercial nuclear plant to support the Radiation Protection Program required in Sec. 53.850.

(b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with Sec. 53.850.

(c) Functional design criteria, including design objectives for dose to the maximally exposed member of the public, must be defined for design features to show that plant design features and corresponding programmatic controls, including monitoring programs, control liquid, gaseous, and solid wastes, as required under part 20 of this chapter.

Sec. 53.430 Design features and functional design criteria for protection of plant workers.

(a) Design features must be provided for each commercial nuclear plant such that, when combined with corresponding programmatic controls, the requirements in Sec. 53.270 can be met.

(b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with Sec. 53.270.

Sec. 53.440 Design requirements.

(a)(1) Analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof must demonstrate that each design feature required by Sec. 53.400 meets the defined functional design criteria required by Sec. Sec. 53.410 and 53.420. This demonstration must consider interdependent effects throughout the commercial nuclear plant and the range of conditions under which the design features required by Sec. 53.400 must function throughout the plant's lifetime.

(2) The design processes for SR and NSRSS SSCs under this part must include administrative procedures for evaluating operating, design, and construction experience and for considering applicable important industry experiences in the design of those SSCs.

(b) The design features classified as SR must, wherever applicable, be designed using generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S. Nuclear Regulatory Commission (NRC).

(c) The materials used for each SR and NSRSS SSC must be qualified for their service conditions over the design life of the SSC as appropriate to satisfy the special treatments established for the SSC under Sec. 53.460.

(d) Possible degradation mechanisms related to aging, fatigue, chemical interactions, operating temperatures, effects of irradiation, and other environmental factors that may affect the performance of SR and NSRSS SSCs must be evaluated and used to inform the design and the development of integrity assessment programs under Sec. 53.870.

(e)(1) Safety-related SSCs and, where appropriate, NSRSS SSCs must be designed and located to minimize, consistent with other safety requirements in this part, the probability and effect of fires and explosions.

(2) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs.

(3) Fire detection and fire suppression systems of appropriate capacity and capability must be provided and designed to minimize the adverse effects of fires on SR and NSRSS SSCs.

(4) Fire suppression systems must be designed to ensure that their rupture or inadvertent operation does not significantly impair the ability of SR and NSRSS SSCs to perform their safety functions to satisfy Sec. 53.230.

(f) Safety and security must be considered together in the design process such that, where possible, security issues are effectively resolved through design and engineered security features.

(g) The reactor system and waste stores for each commercial nuclear plant must be capable of achieving and maintaining a subcritical condition during normal operations and following any LBE identified in accordance with Sec. 53.240.

(h) Each commercial nuclear plant must have a capability to provide long-term cooling of the reactor fuel and waste stores during normal operations and following any LBE identified in accordance with Sec. 53.240.

(i) The design, analysis, staffing, and programmatic controls for each commercial nuclear plant must consider the number of reactors, waste stores, and other significant inventories of radioactive materials and the associated operating configurations, common systems, system interfaces, and system interactions.

(j) [Reserved]

(k) Design features, related functional design criteria, programmatic controls, or a combination thereof must be defined such that analyses demonstrate a low risk of permanent injury to the public due to the health effects of the chemical hazards of licensed material.

(l) Measures must be taken during the design of commercial nuclear plants to minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the generation of radioactive waste in accordance with Sec. 20.1406 of this chapter.

(m)(1) Each commercial nuclear plant must include criticality monitoring capabilities meeting the requirements of either Sec. 70.24 of this chapter or paragraph (m)(2) of this section.

(2) In lieu of maintaining a monitoring system capable of detecting criticality as described in Sec. 70.24 of this chapter, criticality accident requirements may be satisfied by--

(i) Demonstrating the sub-criticality of special nuclear material, except when it is inside the reactor and the reactor is being operated, by maintaining k-effective below 0.95 at a 95 percent probability, 95 percent confidence level, under conditions that maximize reactivity for the applicable storage and handling configurations, and

(ii) Providing radiation monitors for fuel storage and associated handling areas when fuel is present to detect excessive radiation levels and to support initiating appropriate safety actions.

(3) While a spent fuel transportation package approved under 10 CFR part 71 of this chapter or spent fuel storage cask approved under 10 CFR part 72 is in the special nuclear material handing or storage area, the requirements in 10 CFR parts 71 or 72, as applicable, and the requirements of the certificate of compliance for that package or cask, are the applicable requirements for the fuel within that package or cask.

(n)(1) The design of each commercial nuclear plant must reflect state-of-the-art human factors principles for safe and reliable performance in all locations that human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions.

(2) The design must provide for the capabilities described in Sec. 53.730(b) to ensure the plant staff are able to monitor plant conditions and respond to events.

(3) The means by which the design and human actions together will achieve the safety requirements of subpart B of this part must be evaluated and used to inform the design and the development of the concept of operations required by Sec. 53.730(c).

(4) A functional requirements analysis and function allocation must be used to ensure that plant design features address how safety functions and functional safety criteria are satisfied, and how the safety functions will be assigned to appropriate combinations of human action, automation, active safety features, passive safety features, or inherent safety characteristics.

Sec. 53.450 Analysis requirements.

(a) Requirement to have a probabilistic risk assessment (PRA), or other systematic risk evaluations (SREs), or a combination thereof. A PRA, other SREs, or a combination thereof for each commercial nuclear plant must be performed and used together with other generally accepted approaches for systematically evaluating engineered systems to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in Sec. 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of Sec. 53.220.

(b) Specific uses of analyses. The PRA, other SREs, or a combination thereof, together with other generally accepted approaches for systematically evaluating engineered systems must be used to--

(1) Inform the selection of the LBEs, as described in Sec. 53.240, which must be

considered in the design to determine compliance with the safety criteria in subpart B of this part.

(2) Inform the classification of SSCs according to their safety significance in accordance with Sec. 53.460 and to identify the environmental conditions under which the SSCs and operating staff must perform their safety functions.

(3) Evaluate the adequacy of defense-in-depth measures required in accordance with Sec. 53.250.

(4) Identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment.

(5) Identify and assess events that challenge plant control and safety systems whose failure could lead to the uncontrolled release of radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events identified in accordance with subpart D of this part.

(6) Inform the establishment and updating of appropriate measures for plant operations, including availability controls, to ensure that the configurations and special treatments for SR SSCs and NSRSS SSCs provide the capabilities, availability, and reliability consistent with satisfying the safety criteria under Sec. Sec. 53.220 and the analyses of licensing-basis events other than design-basis accidents (DBAs) under Sec. 53.450(e).

(c) Maintenance and upgrade of analyses. The PRA, other SREs, or a combination thereof must be maintained (e.g., updated to reflect plant changes such as modifications, procedure changes, or plant performance data) at least every 5 years until the permanent cessation of operations under Sec. 53.1070 and upgraded (e.g., changed in scope or use of new methods) in conformance with generally accepted methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC.

(d) Qualification of analytical codes. The analytical codes used in modeling the physical behavior of plant systems in the analyses of licensing-basis events (including but not limited to thermodynamics, reactor physics, fuel performance, and mechanistic source term codes) must be qualified for the range of conditions for which they are to be used.

(e) Analyses of licensing-basis events other than design-basis accidents. (1) Analyses must be performed for LBEs other than design- basis accidents (DBAs). These LBEs must be identified using insights from a PRA, other SREs, or a combination thereof with other generally accepted approaches for systematically evaluating engineered systems to identify and analyze equipment failures and human errors.

(2) The analysis of LBEs other than DBAs must include definitions of evaluation criteria for each event or specific categories of LBEs to determine the acceptability of the plant response to the challenges posed by internal and external hazards to provide an appropriate level of safety.

(3) The analyses of LBEs other than DBAs must address event sequences from initiation to a defined end state and be used in combination with other engineering analyses to demonstrate that the functional design criteria required by Sec. 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the evaluation criteria defined for each LBE other than DBAs, to satisfy the safety criteria specified in accordance with Sec. 53.220 and provide defense in depth as required by Sec. 53.250.

(4) The methodology used to identify, categorize, and analyze LBEs must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety.

(f) Analysis of design-basis accidents. (1) The analysis of LBEs required by Sec. 53.240 must include analysis of DBAs that address possible challenges to the safety functions identified under Sec. 53.230. The events selected as DBAs must be those that, if not terminated, have the potential for exceeding the safety criteria in Sec. 53.210.

(2) The DBAs selected must be analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the SR SSCs identified under Sec. 53.460 and human actions addressed by the requirements of subpart F of this part are available to perform the safety functions identified in accordance with Sec. 53.230.

(3) The analysis must conservatively demonstrate compliance with the safety criteria in Sec. 53.210.

(g) Other required analyses. Analyses must be performed to assess--

(1) Fire protection. Fire protection measures to demonstrate, through inclusion of fires in the analysis of LBEs or by separate analyses, that a fire or explosion in any plant area would not--

(i) Prevent equipment from fulfilling the safety functions identified in accordance with Sec. 53.230; or

(ii) Challenge the safety criteria in Sec. Sec. 53.210 and 53.220.

(2) [Reserved]

(3) Dose to members of the public. Measures taken under Sec. 53.425, including estimating--

(i) The quantity of each of the principal radionuclides expected to be released annually to unrestricted areas in liquid effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas.

(ii) The quantities of each of the principal radionuclides of the gases, halides, and particulates expected to be released annually to unrestricted areas in gaseous effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas.

(iii) The annual external radiation dose in unrestricted areas and the maximally exposed member of the public in unrestricted areas due to direct radiation from contained radiation sources from the commercial nuclear plant during normal reactor operations.

Sec. 53.460 Safety categorization and special treatments.

(a) Structures, systems, and components must be classified according to their safety significance. The SSC categories must include “Safety-Related,” “Non-Safety-Related but Safety-Significant,” and “Non-Safety-Significant,” as defined in subpart A of this part.

(b) For SR and NSRSS SSCs, the conditions under which they must perform their safety function in Sec. 53.230 must be identified. Special treatments must be established in accordance with this and other subparts to provide confidence that the SSCs will perform under the service conditions and with reliability consistent with the analysis performed under Sec. 53.450 to demonstrate meeting the safety criteria in Sec. Sec. 53.210 and 53.220.

(1) The special treatments for SR SSCs must include meeting the applicable quality assurance requirements from appendix B of part 50 of this chapter.

(2) The special treatments for NSRSS SSCs and special treatments for SR SSCs beyond those required under paragraph (b)(1) of this section may include meeting selected quality assurance requirements from appendix B of part 50 of this chapter when such treatment is needed to address performance requirements, equipment reliability, or uncertainties.

(c) The identification of special treatments for SR and NSRSS SSCs must account for human actions needed to prevent or mitigate LBEs, the need to perform such actions reliably under the postulated environmental conditions, and the role of programs established in accordance with subpart F of this part to provide confidence that those actions

will be performed as assumed in the analysis performed in accordance with Sec. 53.450 to demonstrate meeting the applicable criteria in Sec. Sec. 53.210, 53.220, and 53.450(e).

Sec. 53.470 [Reserved]

Sec. 53.480 Earthquake engineering.

(a) Effects of earthquakes. Structures, systems, and components classified as SR or NSRSS must be able to withstand the effects of earthquakes, commensurate with the safety significance of the SSC, without loss of capability to perform their role in fulfilling the safety functions required by Sec. 53.230.

(b) Definitions. As used in this section--

Design-Basis Ground Motions (DBGMs) are the vibratory ground motions for which certain SSCs must be designed to remain functional.

Operating basis earthquake (OBE) ground motion is the vibratory ground motion for which those features of the commercial nuclear plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional. The OBE ground motion is used in Sec. 53.720.

Response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports.

Surface deformation is the distortion of geologic strata on or near the ground surface that occurs because of tectonic forces that result from earthquakes.

(c) Design considerations--(1) Design-Basis Ground Motions. (i) The DBGMs must be derived from the Site Ground Motion Response Spectra developed in accordance with Sec. 53.510(c), by taking into consideration the functional design criteria of SSCs in accordance with Sec. Sec. 53.410 and 53.420. The horizontal component of the DBGM(s) in the free-field at the foundation level of the structures must be an appropriate response spectrum that is determined based on the risk- significance of SSCs and their safety functions. In view of the limited data available on vibratory ground motion of strong earthquakes, it is acceptable that the design response spectra be smoothed spectra.

(ii) The commercial nuclear plant must be designed so that, if the DBGMs occur, the following SSCs remain functional and within applicable stress, strain, and deformation limits:

(A) Structures, systems, and components for which functional design criteria are established in accordance with Sec. 53.410 or Sec. 53.420; and

(B) Structures, systems, and components classified as SR or NSRSS commensurate with safety significance in accordance with Sec. 53.460.

(iii) In addition to seismic loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of the SR SSCs and, commensurate with safety significance, NSRSS SSCs.

(iv) The design of the commercial nuclear plant must take into account the possible effects of seismic-induced ground disruption, such as fissuring, lateral spreads, differential settlement, liquefaction, and landsliding, on the facility foundations.

(v) The SSCs fulfilling the safety functions required by Sec. 53.230 must be demonstrated through design, testing, or qualification methods to be able to fulfill those safety functions during and after the vibratory ground motion associated with the DBGMs.

(vi) The evaluation of SSCs required by this section to show they are able to function during and after earthquake ground motion should consider, if applicable, soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for inelastic behavior in some of these SSCs during the DBGMs and under the postulated concurrent loads, provided the necessary safety functions are maintained.

(2) OBE Ground Motion. The OBE Ground Motion must be characterized by response spectra. The value of the OBE Ground Motion must be set to one-third or less of the DBGMs response spectra.

(3) [Reserved]

(4) Required seismic instrumentation. Suitable instrumentation must be provided so that the seismic response of commercial nuclear plant SR SSCs or NSRSS SSCs can be evaluated promptly after an earthquake.

(d) Surface deformation. (1) The potential for surface deformation must be taken into account in the design of the commercial nuclear plant by providing reasonable assurance that in the event of deformation, SSCs classified as SR or NSRSS in accordance with Sec. 53.460 will remain functional.

(2) In addition to surface deformation induced loads, the design of SSCs must take into account, commensurate with safety significance, seismic loads and applicable concurrent functional and accident-induced loads.

(3) The design provisions for surface deformation must be based on its postulated occurrence in any direction and azimuth and under any part of the commercial nuclear plant, unless evidence indicates this assumption is not appropriate, and must take into account the estimated rate at which the surface deformation may occur.

(e) Seismically induced floods and water waves and other design conditions. Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to subpart D of this part must be taken into account in the design of the commercial nuclear plant so as to prevent undue risk to the health and safety of the public.

(f) Analysis. The analyses required by Sec. 53.450 must address seismic hazards and related SSC responses in determining that the safety criteria defined in Sec. 53.220 will be met.

(g) Design criteria, human actions, and programmatic controls. Functional design criteria, human actions, and programmatic controls needed to address seismic events must be identified and implemented in accordance with this and other subparts to achieve and maintain the performance of SSCs relied upon to satisfy the safety criteria in Sec. 53.220 and to maintain consistency with analyses required by Sec. 53.450 when accounting for the site-specific frequencies and magnitudes of earthquakes for a commercial nuclear plant.

Sec. 53.500 General siting and siting assessment.

The purpose of this subpart and the specific requirements therein is to ensure that:

(a) The siting of each commercial nuclear plant is supported by assessments of proposed sites such that the design, including design features and programmatic controls corresponding to the site characteristics, satisfies the safety criteria defined in Sec. Sec. 53.210 and 53.220. The siting assessment addresses the site characteristics that might contribute to the initiation, progression, or consequences of licensing-basis events (LBEs) analyzed under Sec. Sec. 53.450 and 53.480 that are identified and mitigated by design features or programmatic controls. The siting assessment takes into consideration the potential adverse impacts that a commercial nuclear plant may have on nearby populations as a result of normal operations or LBEs.

(b) Activities performed to identify site characteristics or otherwise needed to determine site-specific contributors to

functional design criteria or analysis assumptions under subpart C of this part satisfy the applicable special treatment requirements of Sec. 53.460, including, where applicable, the quality assurance requirements from appendix B of part 50 of this chapter.

Sec. 53.510 External hazards.

(a) General external hazard requirements. The design-basis external hazard level for the relevant external hazards for a site must be identified and characterized based on site-specific assessments of natural and constructed hazards with the potential to adversely affect plant functions. The external hazard frequencies and magnitudes determined from the site-specific assessments must take into account uncertainties and variabilities in data, models, and methods relied on to characterize the external hazards.

(b) Definitions. As used in this section, the following terms mean:

Geological Siting Factors are geological and seismic factors that may affect the design and operation of the proposed commercial nuclear plant.

Ground Motion Response Spectra (GMRS) are the site-specific GMRS resulting from the geologic investigations and evaluations of the site vicinity and region and used to determine design-basis ground motions for structures, systems, and components under Sec. 53.480.

Probabilistic Seismic Hazard Analysis is an analytical methodology that incorporates uncertainty into estimates of an annual frequency of exceedance for a certain ground motion parameter (e.g., peak ground acceleration, peak ground velocity, response spectral values) at a site.

(c) Geological investigations. The GMRS for the site must be determined based on the results of investigations of the geological, seismological, and engineering characteristics of the site and its environs and must be characterized by both horizontal and vertical free-field GMRS at the free ground surface. The size of the region to be investigated and the type of data pertinent to the investigations must be determined based on the nature of the region surrounding the site. Data on vibratory ground motion, earthquake recurrence rates, fault geometry and slip rates, and site subsurface material properties must be obtained by reviewing pertinent literature and carrying out field investigations. Uncertainties are inherent in the parameters and models used to estimate the GMRS for the site. The site assessment must reflect these uncertainties through an appropriate analysis, such as a probabilistic seismic hazard analysis.

(d) Geologic and seismic siting factors. The geologic and seismic siting factors considered for design under Sec. Sec. 53.415 and 53.480 must include, but are not limited to, determination of the potential for surface tectonic and nontectonic deformations, the size and character of seismically induced floods and water waves that could affect a site from either locally or distantly generated seismic activity, soil and rock stability, liquefaction potential, and natural and artificial slope stability.

Sec. 53.520 Site characteristics.

Site characteristics that might contribute to the initiation, progression, or consequences of LBEs analyzed under Sec. 53.450 must be identified, assessed, and considered in the design and analyses required by subpart C of this part.

Sec. 53.530 Population-related considerations.

Every site must have an exclusion area, a low-population zone, and a population center distance as defined in Sec. 53.020.

(a) The offsite radiological consequences estimated by the analyses required by Sec. 53.450(f) must be used to confirm that--

(1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following onset of the postulated fission product release would not receive a radiation dose in excess of 25 rem (250 millisieverts) total effective dose equivalent.

(2) An individual located at any point on the outer boundary of the low-population zone who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem (250 millisieverts) total effective dose equivalent.

(b) The reactor site must either:

(1) Provide a population center distance of at least one and one- third times the distance from the reactor to the outer boundary of the low-population zone; or

(2) Be found acceptable to the U.S. Nuclear Regulatory Commission (NRC) based on assessments of societal risks in comparison to societal benefits for the specific site. The boundary of the population center or the alternate area assessed considering societal risks and benefits must be determined upon consideration of population distribution. Political boundaries are not controlling in the calculation of population center distance or the alternate area assessed considering societal risks and benefits.

(c) Reactor sites should be located away from very densely populated centers or otherwise be shown to be acceptable by assessments of societal risks in comparison to societal benefits for the specific site. Areas of low-population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low- population density or when assessing a site considering societal risks and benefits, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable.

Sec. 53.540 Siting interfaces.

Site characteristics must be addressed by the design features, programmatic controls, and supporting analyses used to demonstrate that the safety criteria in Sec. Sec. 53.210 and 53.220 are met for each commercial nuclear plant. Site characteristics must be such that adequate emergency plans and security plans can be developed and maintained.

Sec. 53.600 Construction and manufacturing--scope and purpose.

This subpart applies to those construction and manufacturing activities authorized by a construction permit (CP), combined license (COL), manufacturing license (ML), or limited work authorization (LWA) issued under this part.

Sec. 53.605 Reporting of defects and noncompliance.

Each CP and ML issued under this part is subject to the terms and conditions in this section, and each COL issued under this part is subject to the terms and conditions in this section until the date that the Commission makes the finding under Sec. 53.1452(g).

(a) Definitions. The definitions in Sec. 21.3 of this chapter apply to this section.

(b) Posting requirements. (1) Each individual, partnership, corporation, dedicating entity, or other entity subject to the regulations in this section must post current copies of this section and the regulations in 10 CFR part 21; section 206 of the Energy Reorganization Act of 1974, as amended; and procedures adopted under these regulations. These documents must be posted in a conspicuous position on any premises within the United States where the

activities subject to the license are conducted.

(2) If posting of these regulations or the procedures adopted under them is not practical, the licensee may, in addition to posting section 206 of the Energy Reorganization Act of 1974, as amended, post a notice that describes the regulations/procedures, including the name of the individual to whom reports may be made, and states where they may be examined.

(c) Procedures. The holder of a CP, COL, or ML subject to this section must adopt appropriate procedures to--

(1) Evaluate deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazards as soon as practicable, and, except as provided in paragraph (c)(2) of this section, in all cases within 60 days of discovery, to identify a reportable defect or failure to comply that could create a substantial safety hazard, were it to remain uncorrected.

(2) Ensure that if an evaluation of an identified deviation or failure to comply potentially associated with a substantial safety hazard cannot be completed within 60 days from the discovery of the deviation or failure to comply, an interim report is prepared and submitted to the Commission through a director or responsible officer, or designated person as discussed in paragraph (d)(5) of this section. The interim report should describe the deviation or failure to comply that is being evaluated and should also state when the evaluation will be completed. This interim report must be submitted in writing within 60 days of discovery of the deviation or failure to comply.

(3) Ensure that a director or responsible officer of the holder of a CP, COL, or ML subject to this section is informed as soon as practicable, and, in all cases, within the 5 working days after completion of the evaluation described in paragraph (c)(1) or (c)(2) of this section, if the construction or manufacture of a facility or activity, or a basic component supplied for such a facility or activity--

(i) Fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard;

(ii) Contains a defect; or

(iii) Underwent any significant breakdown in any portion of the quality assurance program (QAP) conducted under the requirements of appendix B to part 50 of this chapter that could have produced a defect in a basic component. These breakdowns in the QAP are reportable whether or not the breakdown actually resulted in a defect in a design approved and released for construction, installation, or manufacture.

(d) Reporting defects and noncompliance. (1) The holder of a CP, COL, or ML subject to this section that obtains information reasonably indicating that the facility or manufactured reactors fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard must notify the Commission of the failure to comply through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section.

(2) The holder of a CP, COL, or ML subject to this section that obtains information reasonably indicating the existence of any defect found in the construction or manufacture, or any defect found in the final design of a facility as approved and released for construction or manufacture, must notify the Commission of the defect through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section.

(3) The holder of a CP, COL, or ML subject to this part, who obtains information reasonably indicating that the QAP has undergone any significant breakdown discussed in paragraph (c)(3)(iii) of this section must notify the Commission of the breakdown in the QAP through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section.

(4) When acting as a dedicating entity, the holder of a CP, COL, or ML subject to this section is responsible for identifying and evaluating deviations; reporting defects and failures to comply associated with substantial safety hazards for dedicated items; and maintaining auditable records for the dedication process.

(5) The notification requirements of this paragraph (d) apply to all defects and failures to comply associated with a substantial safety hazard regardless of whether extensive evaluation, redesign, or repair is required to conform to the criteria and bases stated in the Safety Analysis Report, CP, COL, or ML. Evaluation of potential defects and failures to comply and reporting of defects and failures to comply under this section satisfies the CP holder's, COL holder's, and ML holder's evaluation and notification obligations under 10 CFR part 21, and satisfies the responsibility of individual directors or responsible officers or holders of a CP, COL, or ML subject to this section to report defects, and failures to comply associated with substantial safety hazards under section 206 of the Energy Reorganization Act of 1974, as amended. The director or responsible officer may authorize an individual to provide the notification required by this section. However, this does not relieve the director or responsible officer of his or her responsibility under this section.

(e) Notification--timing and where sent. The notification required by paragraph (d) of this section must consist of--

(1) Initial notification by telephone, facsimile, or email identified in appendix A to 10 CFR part 73 to the U.S. Nuclear Regulatory Commission (NRC) Operations Center within 2 days following receipt of information by the director or responsible corporate officer under paragraph (c)(3) of this section, on the identification of a defect or a failure to comply. If the CP, COL, or ML holder elects to use facsimile, verification that the facsimile has been received should be made by calling the NRC Operations Center. This paragraph (e)(1) does not apply to interim reports described in paragraph (c)(2) of this section.

(2) Written notification submitted to the NRC Document Control Desk by an appropriate method listed in Sec. 53.040, with a copy to the appropriate NRC Regional Administrator at the address specified in appendix D to 10 CFR part 20 and a copy to the appropriate NRC resident inspector, if applicable, within 30 days following receipt of information by the director or responsible corporate officer under paragraph (c)(3) of this section, on the identification of a defect or failure to comply.

(f) Content of notification. The written notification required by paragraph (e)(2) of this section must clearly indicate that the written notification is being submitted under this section and include the following information, to the extent known.

(1) Name and address of the individual or individuals informing the Commission.

(2) Identification of the facility, the activity, or the basic component supplied for the facility or the activity within the United States which contains a defect or fails to comply.

(3) Identification of the firm constructing or manufacturing the facility or supplying the basic component which fails to comply or contains a defect.

(4) Nature of the defect or failure to comply and the safety hazard which is created or could be created by the defect or failure to comply.

(5) The date on which the information of a defect or failure to comply was obtained.

(6) In the case of a basic component that contains a defect or failure to comply, the number and location of these components in use at the facility subject to the regulations in this part.

(7) In the case of a completed reactor manufactured under this part, the entities to which the reactor was supplied.

(8) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

(9) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to other entities.

(g) Procurement documents. Each holder of a CP, COL, or ML subject to this section must ensure that each procurement document for a facility or a basic component specifies the provisions of 10 CFR part 21 or this section that apply, as applicable.

(h) Coordination with 10 CFR part 21. The requirements of this section are satisfied when the defect or failure to comply associated with a substantial safety hazard has been previously reported under 10 CFR part 21, under Sec. 73.1205 of this chapter, under this section, or under Sec. 53.1640.

(i) Records retention. The holder of a CP, COL, or ML subject to this section must prepare and maintain records necessary to accomplish the purposes of this section, specifically--

(1) Retain procurement documents, which define the requirements that facilities or basic components must satisfy in order to be considered acceptable, for the lifetime of the facility or basic component.

(2) Retain records of evaluations of all deviations and failures to comply under paragraph (c)(1) of this section for the longest of--

(i) Ten years from the date of the evaluation;

(ii) Five years from the date that an early site permit is referenced in an application for a COL; or

(iii) Five years from the date of delivery of a manufactured reactor.

(3) Retain records of all interim reports to the Commission made under paragraph (c)(2) of this section, or notifications to the Commission made under paragraph (d) of this section for the minimum time periods stated in paragraph (i)(2) of this section;

(4) Suppliers of basic components must retain records of--

(i) All notifications sent to affected licensees or purchasers under paragraph (d)(4) of this section for a minimum of 10 years following the date of the notification;

(ii) The facilities or other purchasers to whom the basic components or associated services were supplied for a minimum of 15 years from the delivery of the basic component or associated services.

(5) Maintaining reports in accordance with this section satisfies the recordkeeping obligations under 10 CFR part 21 of the entities, including directors or responsible officers thereof, subject to this section.

Sec. 53.610 Construction.

(a) Management and control. Licensees must ensure that the following plans, programs, and organizational units are developed and implemented to manage and control the construction activities:

(1) Programs to ensure that the construction of a commercial nuclear plant supports the eventual compliance with the design and analysis requirements in subpart C of this part.

(2) An organization, headed by qualified personnel, responsible for managing, controlling, and evaluating the adequacy of the construction activities.

(3) Procedures describing the qualifications for personnel in key positions in the licensee's management and control organization and the organizational responsibilities, authority, and interfaces with other parts of the licensee's organization.

(4) Procedures to evaluate the applicability of other national and international construction experience to the planned and ongoing construction activities and to ensure the applicable experience will be provided to those constructing the plant.

(5) A fitness-for-duty program, under 10 CFR part 26.

(6)(i) A QAP meeting the requirements of appendix B of part 50 of this chapter as required by Sec. 53.460(b).

(ii) Appropriate programmatic controls to provide special treatment for non-safety-related but safety-significant structures, systems, and components (SSCs).

(7) A radiation protection program, in accordance with 10 CFR part 20, that includes measures for monitoring the dose to individuals working with radioactive materials brought onto the site, as applicable.

(8) An information security program in accordance with Sec. Sec. 73.21, 73.22, and 73.23 of this chapter, as applicable.

(b) Construction activities. No person may begin the construction of a commercial nuclear plant on a site on which the facility is to be operated under this part until that person has been issued either a CP or COL, an early site permit authorizing activities under Sec. 53.1130, or an LWA under this part.

(1) Licensees must satisfy the following requirements:

(i) As appropriate, considering the types and quantities of radioactive materials being brought onto the site--

(A) The licensee must maintain and follow a special nuclear material (SNM) material control and accounting program, a measurement control program, and other material control procedures that include corresponding record management requirements as required by the provisions of Sec. 70.32 of this chapter. Prior to initial receipt of SNM onsite, the licensee must implement an SNM material control and accounting program in accordance with 10 CFR part 74.

(B) Procedures must be in place to receive, possess, use, and store source, byproduct, and SNM in accordance with applicable portions of 10 CFR parts 30, 40, and 70.

(C) A plant staff training program associated with the receipt of radioactive material must be approved and implemented prior to initial receipt of byproduct, source or SNM (excluding exempt quantities as described in Sec. 30.18 of this chapter).

(ii) For construction of a commercial nuclear plant involving multiple reactor units, plans and procedures must be in place to prevent or mitigate potential hazards to the SSCs of operating units resulting from construction activities, including the managerial and administrative controls to be used to provide assurance that the limiting conditions for operation of the operating units are not exceeded as a result of construction activities.

(iii) Procedures must be in place prior to the start of construction activities that describe how construction will be controlled so as not to impact other features important to the design, such as dewatering, slope stability, backfill, compaction, and seepage.

(iv) For LWA holders, a plan must be developed for redress of activities performed under the LWA should one of the following situations arise:

(A) LWA work activities are terminated by the holder of the LWA;

(B) The LWA is revoked by the NRC; or

(C) The Commission denies the associated CP or COL application.

(2)(i) Onsite fresh fuel must be protected and stored in compliance with Sec. 73.67 of this chapter.

(ii) Before initial fuel load into the reactor (or, for a fueled manufactured reactor, before initiating the removal of the features to prevent criticality required under Sec. 53.620(d)(1)), a cybersecurity program that meets the requirements of Sec. 73.54 or Sec. 73.110 of this chapter, a physical security program that meets the requirements of Sec. 73.55 or Sec. 73.100 of this chapter, and an access authorization program that meets the requirements of Sec. 73.56 or Sec. 73.120 of this chapter must be established, as applicable.

(iii) Fire protection measures must be implemented for work and storage areas (including adjacent fire areas that could affect the work or storage area) before initial receipt of byproduct, source, or non- fuel SNM (excluding exempt quantities as described in Sec. 30.18 of this chapter). The fire protection measures for areas associated with new fuel (including all fuel handling, fuel storage, and adjacent fire areas that could affect the new fuel) must be implemented before receipt of fuel. Prior to the receipt of fuel, a formal letter of agreement must be in place with the local fire department specifying the nature of arrangements in support of the fire protection program.

(c) Inspection and acceptance. (1) The licensee must have a process for accepting individual or groups of SSCs upon completion of construction and protecting them from damage or tampering as other construction activities continue.

(2) The post-construction acceptance process must address the inspections, tests, analyses, and acceptance criteria specified in the COL under Sec. 53.1440 or the equivalent verifications needed to support the issuance of an operating license under Sec. 53.1387.

Sec. 53.620 Manufacturing.

(a) Management and control. Holders of MLs must ensure that the following plans, programs, and organizational units are developed and implemented to manage and control the manufacturing activities within the scope of the ML:

(1) Programs to ensure that the manufacturing of a manufactured reactor or portions of a manufactured reactor complies with the design and analysis requirements in subpart C of this part. The entity with design authority for the manufactured reactor covered by the ML must be identified in the license.

(2) An organizational and management structure responsible for managing, controlling, and evaluating the adequacy of the reactor design and manufacturing activities.

(3) Procedures describing the qualifications for personnel in key positions in the licensee's management and control organization and the organizational responsibilities, authority, and interfaces with other parts of the licensee's organization.

(4) A program to evaluate the applicability of other national and international design and manufacturing experience to the planned and ongoing manufacturing activities.

(5) A fitness-for-duty program, in accordance with 10 CFR part 26.

(6)(i) A QAP meeting the requirements of appendix B to part 50 of this chapter, to be applied to the design, fabrication, construction, and testing of the SSCs of the manufactured reactor.

(ii) Appropriate programmatic controls to provide special treatment measures for non-safety-related but safety-significant SSCs.

(7) A radiation protection program, in accordance with 10 CFR part 20, that includes measures for monitoring the dose to individuals if the manufacturing activities include working with radioactive materials.

(8) An information security program in accordance with Sec. Sec. 73.21, 73.22 and 73.23 of this chapter, as applicable.

(b) Manufacturing activities. Holders of MLs must satisfy the following requirements:

(1) The manufacturing process must be conducted within facilities for which the ML holder has the authority to establish controls on any activity that might affect manufacturing. The licensee must establish access controls to the portions of each facility involved in the manufacturing processes governed by the ML.

(2) Manufacturing processes must be performed in accordance with the ML and the referenced codes and standards that have been endorsed or otherwise found acceptable by the NRC.

(3) A post-manufacturing inspection and acceptance process must be established and implemented before transporting a manufactured reactor or portions of a manufactured reactor for installation at a commercial nuclear plant. The process must consider the results of inspections, tests, and analyses that have been performed and the acceptance criteria that are necessary and sufficient to conclude that manufacturing activities have been completed in accordance with the ML.

(c) Control of radioactive materials. As appropriate considering the types and quantities of radioactive materials being brought into the manufacturing facility--

(1) Procedures must be in place to receive, transfer, possess, and use source, byproduct, and SNM in accordance with the applicable portions of 10 CFR parts 30, 40 and 70.

(2) A fire protection program must be established and implemented before the initial receipt of byproduct, source, or non-fuel SNM (excluding exempt quantities as described in Sec. 30.18 of this chapter).

(3) An emergency plan appropriate for responding to the facility- specific hazards of an accidental release of radioactive material and to limit the health effects of the associated chemical hazards of licensed material must be approved and implemented prior to the receipt of byproduct, source, or SNM (excluding exempt quantities as described in Sec. 30.18 of this chapter).

(4) A plant staff training program associated with the receipt of radioactive material must be approved and implemented before initial receipt of byproduct, source, or SNM (excluding exempt quantities as described in Sec. 30.18 of this chapter).

(5) Security requirements must be implemented for the protection of SNM based on the type, enrichment, and quantity in accordance with 10 CFR part 73, as applicable, and for the protection of Category 1 and Category 2 quantities of radioactive material in accordance with 10 CFR part 37, as applicable.

(d) Fuel loading. (1)(i) An ML may authorize possession of a manufactured reactor into which the licensee has loaded fresh (unirradiated) fuel pursuant to a license issued under part 70 of this chapter only if the manufactured reactor is configured during its loading, storage, and transport with features to prevent criticality that are specified in the ML.

(ii) The ML applicant may file a separate, subsequent application for the 10 CFR part 70 license or combine the application for the 10 CFR part 70 license with the application for an ML.

(iii) The Commission has determined that any such fueled manufactured reactor in which the features to prevent criticality are in place is not in operation.

(iv) Upon installation of the fueled manufactured reactor in its place of operation and a Commission finding that the acceptance criteria in the COL that authorized reactor construction are met under Sec. 53.1452(g), or that any conditions in the CP that authorized reactor construction are met and the associated operating license (OL) issued, the features to prevent criticality may be removed. Upon initiating the removal of

the features to prevent criticality, the fueled manufactured reactor has commenced operation.

(2) Holders of part 70 licenses authorizing the possession and loading of fresh fuel into manufactured reactors must comply with the requirements of part 70 for the facilities and activities related to the storage, movement, and loading of fresh fuel in the manufactured reactor. Holders of these part 70 licenses must comply with the requirements of Subpart H to part 70, regardless of whether their proposed activities meet the applicability criteria found in 10 CFR 70.60. Procedures, equipment, and personnel required by the 10 CFR part 70 license, must be in place before the receipt of SNM at the manufacturing facility.

(i) Before the receipt of SNM, the licensee must have security programs in place that meet the performance objectives of 10 CFR 73.67, with the following additions and exceptions:

(A) A physical security plan describing the physical security program must be maintained and a cybersecurity program must be established for the possession and loading of fresh fuel into a manufactured reactor authorized by a 10 CFR part 70 license, regardless of fuel type, enrichment, and quantity.

(B) The physical security program must be designed to prevent unintended and uncontrolled criticality events.

(C) The cybersecurity program must provide reasonable assurance that a cyberattack does not adversely impact the functions performed by digital assets necessary for implementing the physical security requirements of this section, or the radiation monitoring and criticality requirements in this section or in 10 CFR part 70.

(D) All holders of a part 70 license that authorizes loading of fresh fuel into a manufactured reactor must perform the screening required in Sec. 73.67(d)(4) of this chapter to confirm the identity, trustworthiness, and reliability of individuals prior to granting unescorted access to special nuclear material; these determinations must be documented.

(ii) [Reserved]

(3) The loading or unloading of fresh fuel into or from a manufactured reactor and any changes to the configuration of reactivity control and prevention systems for the fueled manufactured reactor must be performed by a certified fuel handler meeting the requirements in subpart F of this part.

(e) Transportation. (1) A holder of an ML may not transport or allow to be removed from the places of manufacture the manufactured reactor or portions thereof as defined in the ML except for either transport to a site for which the Commission has issued a COL or CP that references the subject ML or export in accordance with 10 CFR part 110.

(2) A holder of an ML must include in any contract governing the transport of a manufactured reactor or portions thereof as defined in the ML from the places of manufacture to any other location, a provision requiring that the person transporting the manufactured reactor comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses.

(3) Procedures governing the preparation of the manufactured reactor or portions thereof as defined in the ML for transport and the conduct of the transport must be issued prior to transport. The procedures must implement the protective measures and restrictions described in NRC regulations and NRC-issued licenses to protect the reactor from potential conditions that would adversely affect the safe operation of a commercial nuclear plant.

(4) For a manufactured reactor that is to be loaded with fresh fuel before transport to the place of operation, the ML must specify that transportation will be in accordance with parts 71 and 73 of this chapter.

(f) Acceptance and installation at the site for which the Commission has issued a COL or CP that references the subject ML. (1) Installation at the site for which the Commission has issued a COL or CP that references the subject ML must follow the regulations in Sec. 53.610.

(2) Upon arrival at the site, the manufactured reactor or portions of a manufactured reactor may not be installed in its place of operation unless the COL or CP holder performs inspections sufficient to verify the reactor is in compliance with the ML and has not been damaged in transit. The COL or CP holder must perform these inspections in accordance with documented procedures subject to quality assurance measures commensurate with their importance to safety. In addition, inspections must confirm that the interface requirements between the manufactured reactor or portions of a manufactured reactor and the remaining portions of the commercial nuclear plant are met.