PART 53 [RESERVED]

• Citation: 10 CFR Part 53
• Sections: 0

§ 53.015 Scope.

Subpart A provides general provisions applicable to all applicants and licensees subject to the rules of this part.

§ 53.020 Definitions.

As used in this part:

Anticipated event sequence means event sequences expected to occur one or more times during the life of a commercial nuclear plant. Anticipated event sequences take into account the expected response of all structures, systems, and components (SSCs) within the plant, regardless of safety classification.

Applicant means a person applying for a license, permit, or other form of Commission permission or approval under this part.

Certified fuel handler means, for a commercial nuclear plant, either—

(1) A non-licensed operator who has qualified in accordance with a fuel handler training program approved by the Commission; or

(2) A non-licensed operator who demonstrates compliance with the following criteria:

(i) Has qualified in accordance with a fuel handler training program that demonstrates compliance with the same requirements as training programs for non-licensed operators required by § 53.830, and

(ii) Is responsible for decisions on—

(A) Safe conduct of decommissioning activities,

(B) Safe handling and storage of spent fuel; and

(C) Appropriate response to plant emergencies.

Combined license (COL) means a combined construction permit (CP) and operating license (OL) with conditions for a commercial nuclear plant issued under this part.

Commercial nuclear plant means a facility consisting of one or more commercial nuclear reactors and associated co-located support facilities, including the collection of buildings, radionuclide sources, and SSCs for which a license, certification, or approval is being sought under this part, that is or will be used for producing power for commercial electric power or other commercial purposes. For the purposes of requirements in this part that reference requirements in part 50 of this chapter, a commercial nuclear plant is equivalent to a nuclear power plant.

Commercial nuclear reactor means an apparatus, other than an atomic weapon, designed or used to sustain nuclear fission. For the purposes of requirements in this part that reference requirements in 10 CFR part 50, a commercial nuclear reactor is equivalent to a nuclear reactor as defined in § 50.2 of this chapter.

Commission means the U.S. Nuclear Regulatory Commission (NRC) or its duly authorized representatives.

Construction means the activities in paragraph (1) of this definition and does not mean the activities in paragraph (2) of this definition.

(1) Activities constituting construction are those activities that are conducted on-site to build the commercial nuclear plant, including the driving of piles; subsurface preparation; placement of backfill, concrete, or permanent retaining walls within an excavation; installation of foundations; or in-place assembly, erection, fabrication, or testing, which are for—

(i) Safety-related (SR) SSCs and those non-safety-related but safety-significant (NSRSS) SSCs of a facility for which special treatment includes requirements on design or installation, including associated quality assurance measures;

(ii) SSCs necessary to comply with 10 CFR part 73; or

(iii) Onsite emergency facilities necessary to comply with § 53.855.

(2) Construction does not include—

(i) Changes for temporary use of the land for public recreational purposes;

(ii) Site exploration, including necessary borings to determine foundation conditions or other preconstruction monitoring to establish background information related to the suitability of the site, the environmental impacts of construction or operation, or the protection of environmental values;

(iii) Preparation of a site for construction of a facility, including clearing of the site, grading, installation of drainage, erosion, and other environmental mitigation measures, and construction of temporary roads and borrow areas;

(iv) Erection of fences and other access control measures;

(v) Excavation;

(vi) Erection of support buildings (such as construction equipment storage sheds, warehouse and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and office buildings) for use in connection with the construction of the facility;

(vii) Building of service facilities (such as paved roads, parking lots, railroad spurs, exterior utility and lighting systems, potable water systems, sanitary sewage treatment facilities, and transmission lines);

(viii) Procurement or fabrication of components or portions of the proposed facility occurring at locations other than the final, in-place location at the facility; or

(ix) Manufacture of a nuclear power reactor under a manufacturing license (ML) under subpart H of this part to be installed at the proposed site and to be part of the proposed facility.

Custom combined license (custom COL) means a COL that does not reference a standard design approval, standard design certification, or manufacturing license.

Decommission or decommissioning means to remove a plant or site safely from service and reduce residual radioactivity to a level that permits—

(1) Release of the property for unrestricted use and termination of the license; or

(2) Release of the property under restricted conditions and termination of the license.

Defense in depth means inclusion of two or more independent and redundant layers of defense in the design of a facility and its operating procedures to compensate for uncertainties such that no single layer of defense, no matter how robust, is exclusively relied upon. Defense in depth includes, but is not limited to, the use of access controls, physical barriers, redundant and diverse safety functions, and emergency response measures.

Design-basis accidents (DBAs) means postulated event sequences that are used to set functional design criteria and performance objectives for the design of SR SSCs through deterministic analyses. Design-basis accidents are a type of licensing-basis event and are based on the capabilities and reliabilities of SR SSCs needed to mitigate and prevent event sequences, respectively.

Design-basis external hazard level means the level of severity or intensity of an external hazard for which the SR SSCs are protected against or designed to withstand without losing their capability to perform their safety functions.

Design features means the active and passive SSCs and the inherent characteristics of those SSCs that contribute to limiting the total effective dose equivalent to individual members of the public during normal operations and prevent or mitigate the consequences of event sequences.

Early site permit (ESP) means a Commission approval, issued under subpart H of this part, for a site for one or more commercial nuclear plants. An early site permit is a partial construction permit.

Electric utility means any entity that generates or distributes electricity and that recovers the cost of this electricity, either directly or indirectly, through rates established by the entity itself or by a separate regulatory authority. Investor-owned utilities, including generation or distribution subsidiaries, public utility districts, municipalities, rural electric cooperatives, and State and Federal agencies, including associations of any of the foregoing, are included within the meaning of “electric utility.”

Event sequence means a postulated initiating event defined for a set of initial plant conditions followed by system, safety function, and operator successes or failures, and terminating in a specified end state depending on the system, safety function, and operator successes and failures ( e.g., prevention of release of radioactive material or release in one of the reactor-specific release categories). An event sequence may include many unique variations of events that are similar in terms of results or end states.

Exclusion area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area must normally be prohibited. In any event, residents must be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.

Fission product release means the amount and composition of radioactive material released to the environment, after accounting for any retention of radionuclides provided by reactor design features.

Fuel means special nuclear material (SNM) or source material, discrete elements that physically contain SNM or source material, and homogeneous mixtures that contain SNM or source material, intended to or used to create power in a commercial nuclear plant.

Functional design criteria means metrics for the performance of SSCs. For SR SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.210. For NSRSS SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.220.

License, when used in the context of a facility, means a limited work authorization, CP, OL, early site permit, COL, or ML under this part, or a renewed license issued by the Commission under this part. When used in the context of a license authorizing an individual to manipulate the controls of a facility, license means a license issued by the Commission to perform the function of an operator, senior operator, or generally licensed reactor operator as defined in this part.

Licensee means a person who is authorized to conduct activities under a license issued under this part by the Commission.

Licensing-basis events means a collection of event sequences considered in the design and licensing of the commercial nuclear plant. Licensing-basis events are unplanned events and include anticipated event sequences, unlikely event sequences, very unlikely event sequences, and DBAs.

Licensing-basis information means the information contained in regulations, orders, licenses, certifications, or approvals issued by the NRC for a commercial nuclear plant licensed under this part and that information submitted to the NRC by an applicant or licensee in a Safety Analysis Report, program description, or other licensing-related document required under this part.

Low-population zone means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken on their behalf in the event of a serious accident. A permissible population density or total population within this zone is not included in this definition because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated from a specific area or instructed to take shelter on a timely basis, will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area.

Major decommissioning activity means, for a commercial nuclear plant, any activity that results in permanent removal of major radioactive components, permanently modifies the structure of the containment, if applicable, or results in dismantling components for shipment containing greater than class C waste in accordance with § 61.55 of this chapter.

Major feature of the emergency plans means an aspect of those plans necessary to:

(1) Address in whole or part either one or more of the 16 standards in 10 CFR 50.47(b) or the requirements of 10 CFR 50.160(b), as applicable; or

(2) Describe the emergency planning zones as required in 10 CFR 53.1109(g).

Manufactured reactor means the essential portions of a nuclear reactor that are manufactured under an ML and subsequently transported and incorporated into a commercial nuclear plant under a COL or CP.

Manufacturing license means a license issued under this part that authorizes the manufacture of manufactured reactors but not its construction, installation, or operation.

Non-Safety-Related but Safety-Significant (NSRSS) SSCs means those SSCs which are not SR but are relied on to achieve adequate defense in depth or perform risk-significant functions and warrant special treatment.

Non-Safety-Significant SSCs means those SSCs that are not SR or NSRSS, are not relied on to achieve adequate defense in depth or to perform risk-significant functions, and do not warrant special treatment.

Person means—

(1) Any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, government agency other than the Commission or the Department of Energy, except that the Department of Energy shall be considered a person to the extent that its facilities are subject to the licensing and related regulatory authority of the Commission pursuant to section 202 of the Energy Reorganization Act of 1974, any State or any political subdivision of, or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and

(2) Any legal successor, representative, agent, or agency of the foregoing.

Population center distance means the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents.

Programmatic controls means administrative measures that govern human action in implementing programs and operating, monitoring, and maintaining SSCs and equipment of a commercial nuclear plant. Programmatic controls considered to be licensing basis information are addressed by programs under § 53.845 and are specified in an application for a requested activity of the Commission.

Quality assurance (QA) means all those planned and systematic actions necessary to ensure that a structure, system, or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those QA actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements.

Safety criteria means performance-based metrics that establish a level of safety provided in requirements in §§ 53.210 and 53.220.

Safety-related structures, systems, or components means those SSCs that are relied upon to demonstrate compliance with the safety criteria in § 53.210 and warrant special treatment.

Small modular reactor means a power reactor, which may be of modular design as defined in § 52.1 of this chapter, licensed under this part to produce heat energy up to 1,000 megawatts thermal per module.

Site characteristics means the actual physical, environmental, and demographic features of a site. Site characteristics are specified in an early site permit or in a Preliminary or Final Safety Analysis Report for a limited work authorization, CP, or COL, as applicable.

Site parameters are the postulated physical, environmental, and demographic features of an assumed site. Site parameters are specified in a standard design approval, standard design certification, or ML.

Source material means source material as defined in subsection 11z. of the Atomic Energy Act of 1954, as amended, (the Act) and in the regulations contained in part 40 of this chapter.

Special nuclear material (SNM) means:

(1) Plutonium, uranium-233, uranium enriched in the isotope-233 or in the isotope-235, and any other material which the Commission, pursuant to the provisions of section 51 of the Act, determines to be SNM, but does not include source material; or

(2) Any material artificially enriched by any of the foregoing, but does not include source material.

Special treatment means those requirements, such as QA, design criteria, and programmatic controls, that are taken beyond the procurement, installation, and maintenance of commercial grade products to ensure that SR and NSRSS SSCs will provide defense in depth or perform risk-significant functions. The requirements also ensure that the SSCs will perform under the service conditions and with the reliability assumed in the analysis performed under § 53.450 to demonstrate compliance with the safety criteria in §§ 53.210 for SR SSCs and 53.220 for SR and NSRSS SSCs.

Standard design means a design which is sufficiently detailed and complete to support certification or approval in accordance with subpart H of this part, and which is usable under of this part for a multiple number of units or at a multiple number of sites without reopening or repeating the review.

Standard design approval or design approval means an NRC staff approval, issued under subpart H of this part, of a final standard design for a commercial nuclear plant. The approval may be for either the final design for the entire reactor facility or the final design of major portions thereof.

Standard design certification or design certification means a Commission approval, issued under subpart H of this part, of a final standard design for a nuclear power facility. This design may be referred to as a certified standard design.

Total effective dose equivalent means the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

Utilization facility means any commercial nuclear reactor other than one designed or used primarily for the formation of plutonium or uranium-233.

Unlikely event sequences means event sequences that are not expected to occur in the life of a commercial nuclear plant and are less likely than anticipated event sequences, but are infrequent rather than rare. Unlikely event sequences take into account the expected response of all SSCs within the plant regardless of safety classification.

Very unlikely event sequences means event sequences that are not expected to occur in the life of a commercial nuclear plant, are less likely than an unlikely event sequence, and are rare. Very unlikely event sequences take into account the expected response of all SSCs within the plant regardless of safety classification.

§ 53.040 Written communications.

(a) General requirements. All correspondence, reports, applications, and other written communications from the applicant or licensee to the NRC concerning the regulations in this part or individual license conditions must be sent either by mail addressed: ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, between the hours of 8:15 a.m. and 4 p.m. eastern time; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, email, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's website at https://www.nrc.gov/site-help/e-submittals.html; by email to [email protected]; or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the communication is on paper, the signed original must be sent. If a submission due date falls on a Saturday, Sunday, or Federal holiday, the next Federal working day becomes the official due date.

(b) Distribution requirements. Copies of all correspondence, reports, and other written communications concerning the regulations in this part or individual license conditions, or the terms and conditions of an early site permit or standard design approval, must be submitted to the persons listed below (addresses for the NRC Regional Offices are listed in appendix D to 10 CFR part 20).

(1) Applications for amendment of permits and licenses, reports, and other communications. All written communications (including responses to generic letters, bulletins, information notices, regulatory information summaries, inspection reports, and miscellaneous requests for additional information) that are required of holders of licenses, permits, and design approvals issued pursuant to this part, must be submitted as follows, except as otherwise specified in paragraphs (b)(2) through (7) of this section: to the NRC's Document Control Desk (if on paper, the signed original), with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part.

(2) Applications for permits and licenses, and amendments to applications. Applications for licenses, permits, and design approvals and amendments to any of these types of applications must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the facility or the place of manufacture of a reactor licensed under this part, except as otherwise specified in paragraphs (b)(3) through (9) of this section. If the application or amendment is on paper, the submission to the Document Control Desk must be the signed original.

(3) Acceptance review application. Written communications required for an application for determination of suitability for docketing must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original.

(4) Security plan and related submissions. Written communications, as defined in paragraphs (b)(4)(i) through (v) of this section, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original. Submissions should include the following as appropriate:

(i) Physical security plan;

(ii) Safeguards contingency plan;

(iii) Cybersecurity plan;

(iv) Change to security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan made without prior Commission approval under § 53.1565; and

(v) Application for amendment of physical security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan under § 53.1510.

(5) Emergency plan and related submissions. Written communications as defined in paragraphs (b)(5)(i) through (iii) of this section must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility. If the communication is on paper, the submission to the Document Control Desk must be the signed original. Submissions should include the following as appropriate:

(i) Emergency plan;

(ii) Change to an emergency plan under § 53.1565; and

(iii) Emergency implementing procedures under § 53.855.

(6) Updated Final Safety Analysis Report. An updated Final Safety Analysis Report or replacement pages under § 53.1545 must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part. Paper copy submissions may be made using replacement pages; however, if a licensee chooses to use electronic submission, all subsequent updates or submissions must be performed electronically on a total replacement basis. If the communication is on paper, the submission to the Document Control Desk must be the signed original. If the communications are submitted electronically, see Guidance for Electronic Submissions to the Commission.

(7) Quality assurance related submissions. (i) A change to the Safety Analysis Report QA program description under § 53.1565, or a change to a licensee's NRC-accepted QA topical report under § 53.1565, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part. If the communication is on paper, the submission to the Document Control Desk must be the signed original.

(ii) A change to an NRC-accepted QA topical report from non-licensees ( i.e., architect/engineers, nuclear steam supply system suppliers, fuel suppliers, constructors, etc.) must be submitted to the NRC's Document Control Desk. If the communication is on paper, the signed original must be sent.

(8) Certification of permanent cessation of operations. The licensee's certification of permanent cessation of operations, under subpart G of this part, must state the date on which operations have ceased or will cease, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.

(9) Certification of permanent fuel removal. The licensee's certification of permanent fuel removal, under subpart G of this part, must state the date on which the fuel was removed from the reactor vessel and the disposition of the fuel, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation.

(c) Form of communications. All paper copies submitted to demonstrate compliance with the requirements set forth in paragraph (b) of this section must be typewritten, printed, or otherwise reproduced in permanent form on unglazed paper. Exceptions to these requirements imposed on paper submissions may be granted for the submission of micrographic, photographic, or similar forms.

(d) Regulation governing submission. Licensees, applicants, and holders of standard design approvals submitting correspondence, reports, and other written communications under the regulations of this part are requested but not required to cite whenever practical, in the upper right corner of the first page of the submission, the specific regulation or other basis requiring submission.

§ 53.050 Deliberate misconduct.

(a) Any licensee or applicant for a license; holder of or applicant for a standard design approval; applicant for a standard design certification; employee of a licensee, holder of a standard design approval, or applicant for a license, standard design approval, or standard design certification; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, holder of or applicant for a standard design approval, or applicant for a standard design certification, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not—

(1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or

(2) Deliberately submit to the NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.

(b) A person who violates paragraph (a)(1) or (2) of this section may be subject to enforcement action in accordance with the procedures in subpart B of 10 CFR part 2.

(c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows—

(1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or

(2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.

§ 53.060 Employee protection.

(a) Discrimination by a Commission licensee, holder of a standard design approval, an applicant for a license, standard design certification, or standard design approval, a contractor or subcontractor of a Commission licensee, holder of a standard design approval, applicant for a license, standard design certification, or standard design approval, against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Act or the Energy Reorganization Act of 1974, as amended.

(1) The protected activities include but are not limited to—

(i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those statutes;

(ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer;

(iii) Requesting the NRC to institute action against his or her employer for the administration or enforcement of these requirements;

(iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section; and

(v) Assisting or participating in, or being about to assist or participate in, these activities.

(2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.

(3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Act.

(b) Any employee who believes that they have been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.

(c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, a holder of a standard design approval, an applicant for a Commission license, standard design certification, or a standard design approval, or a contractor or subcontractor of a Commission licensee, holder of a standard design approval, or any applicant may be grounds for—

(1) Denial, revocation, or suspension of the license or standard design approval;

(2) Withdrawal or revocation of a proposed or final standard design certification;

(3) Imposition of a civil penalty on the licensee, holder of a standard design approval, or applicant (including an applicant for a standard design certification under this part following Commission adoption of final design certification rule) or a contractor or subcontractor of the licensee, holder of a standard design approval, or applicant; or

(4) Other enforcement action.

(d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.

(e)(1) Each holder or applicant for a license or design approval, must prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in § 19.11(e)(1) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted no later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

(2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate NRC Regional Office listed in appendix D to 10 CFR part 20, via email to [email protected], or by visiting the NRC's online library at https://www.nrc.gov/reading-rm/doc-collections/forms/.

(f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section, including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.

(g) Part 19 of 10 CFR sets forth requirements and regulatory provisions applicable to licensees, holders of a standard design approval, applicants for a license, standard design certification, or standard design approval, and contractors or subcontractors of a Commission licensee, or holder of a standard design approval, and are in addition to the requirements in this section.

§ 53.070 Completeness and accuracy of information.

(a) Information provided to the Commission by a holder of a license, permit, design certification, or standard design approval under this part or an applicant for a license, permit, design certification, or standard design approval under this part, and information required by statute or by the Commission's regulations, orders, license conditions, or terms and conditions of a standard design approval to be maintained by the applicant or the licensee must be complete and accurate in all material respects.

(b) Each applicant or licensee, each holder of a standard design approval under this part, and each applicant for a standard design certification under this part following Commission adoption of a final design certification regulation, must notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant, licensee, or holder violates this paragraph (b) only if the applicant, licensee, or holder fails to notify the Commission of information that the applicant, licensee, or holder has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.

§ 53.080 Specific exemptions.

(a) The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.

(b) The Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever—

(1) Application of the regulation in the particular circumstances conflicts with other rules or requirements of the Commission;

(2) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule;

(3) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated;

(4) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption;

(5) The exemption would provide only temporary relief from the applicable regulation and the licensee or applicant has made good faith efforts to comply with the regulation; or

(6) There is present any other material circumstance not considered when the regulation was adopted for which it would be in the public interest to grant an exemption. If such condition is relied on exclusively for demonstrating compliance with paragraph (b) of this section, the exemption may not be granted until the Executive Director for Operations has consulted with the Commission.

(c) Any person may request an exemption permitting the conduct of construction activities prior to the issuance of a CP. The Commission may grant such an exemption upon considering and balancing the following factors:

(1) Whether conduct of the proposed activities will give rise to a significant adverse impact on the environment and the nature and extent of such impact, if any;

(2) Whether redress of any adverse environment impact from conduct of the proposed activities can reasonably be effective should such redress be necessary;

(3) Whether conduct of the proposed activities would foreclose subsequent adoption of alternatives; and

(4) The effect of delay in conducting such activities on the public interest, including whether the power needs to be used by the proposed facility, the availability of alternative sources, if any, to meet those needs on a timely basis, and delay costs to the applicant and to consumers.

(d) Issuance of such an exemption must not be deemed to constitute a commitment to issue a CP. During the period of any exemption granted pursuant to paragraph (c) of this section, any activities conducted must be carried out in such a manner as will minimize or reduce their environmental impact.

(e) The Commission's consideration of requests for exemptions from requirements of the regulations of other parts in this chapter that are applicable by virtue of this part must be governed by the exemption requirements of those parts.

§ 53.090 Standards for review.

(a) Common standards. In determining that a CP, OL, early site permit, COL, or ML under this part will be issued to an applicant, the Commission will be guided by the following considerations:

(1) Except for an early site permit or ML, the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing, collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in 10 CFR part 20, and that the health and safety of the public will not be endangered.

(2) The applicant for a CP, OL, COL, or ML is technically and financially qualified to engage in the proposed activities in accordance with the regulations in this chapter. However, no consideration of financial qualification is necessary for an electric utility applicant for an OL for a utilization facility of the type described in paragraph (d) of this section or for an applicant for an ML.

(3) The issuance of a CP, OL, early site permit, COL, or ML to the applicant will not, in the opinion of the Commission, be inimical to the common defense and security or to the health and safety of the public.

(4) Any applicable requirements of 10 CFR part 51 have been satisfied.

(b) Additional standards for licenses. In determining whether a license will be issued to an applicant, the Commission will, in addition to applying the standards set forth in paragraph (a) of this section, consider whether the proposed activities will serve a useful purpose proportionate to the quantities of SNM or source material to be utilized.

(c) Additional standards and provisions affecting licenses for commercial power. In addition to applying the standards set forth in paragraphs (a) and (b) of this section, paragraphs (c)(1) through (c)(4) of this section apply in the case of a license for a facility for the generation of commercial power.

(1) The NRC will—

(i) Give notice in writing of each application to the regulatory agency or State as may have jurisdiction over the rates and services incident to the proposed activity;

(ii) Publish notice of the application in trade or news publications as it deems appropriate to give reasonable notice to municipalities, private utilities, public bodies, and cooperatives which might have a potential interest in the utilization or production facility; and

(iii) Publish notice of the application once each week for four consecutive weeks in the Federal Register. No license will be issued by the NRC prior to the giving of these notices and until four weeks after the last notice is published in the Federal Register .

(2) If there are conflicting applications for a limited opportunity for such license, the Commission will give preferred consideration in the following order: first, to applications submitted by public or cooperative bodies for facilities to be located in high cost power areas in the United States; second, to applications submitted by others for facilities to be located in such areas; third, to applications submitted by public or cooperative bodies for facilities to be located in areas other than high cost power areas; and, fourth, to all other applicants.

(3) The licensee who transmits electric energy in interstate commerce, or sells it at wholesale in interstate commerce, must be subject to the regulatory provisions of the Federal Power Act.

(4) Nothing will preclude any government agency, now or hereafter authorized by law to engage in the production, marketing, or distribution of electric energy, if otherwise qualified, from obtaining a CP, OL, or COL under this part for a utilization facility for the primary purpose of producing electric energy for disposition for ultimate public consumption.

(d) Licenses for commercial nuclear plants. A license will be issued, to an applicant who qualifies, for any one or more of the following: to transfer or receive in interstate commerce, or manufacture, produce, transfer, acquire, possess, or use a utilization facility for industrial or commercial purposes.

§ 53.100 Jurisdictional limits.

No permit, license, standard design approval, or standard design certification under this part shall be deemed to have been issued for activities that are not under or within the jurisdiction of the United States.

§ 53.000 Purpose.

This part provides an optional, technology-inclusive, performance-based framework for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under section 103 of the Atomic Energy Act of 1954, as amended (the Act) (68 Stat. 919), and Title II of the Energy Reorganization Act of 1974, as amended (88 Stat. 1242). Also, this part gives notice to all persons who knowingly provide to any holder of or applicant for an approval, certification, permit, or license, or to a contractor, subcontractor, or consultant of any of them, components, equipment, materials, or other goods or services that relate to the activities of a holder of or applicant for an approval, certification, permit, or license, subject to this part, that they may be individually subject to U.S. Nuclear Regulatory Commission enforcement action for violation of the provisions in § 53.050.

§ 53.110 Attacks and destructive acts.

Licensees, applicants for licenses, permits, certifications, and design approvals, and applicants for an amendment to any license, permit, certification, or design approval under this part are not required to provide for design features or other measures for the specific purpose of protection against the effects of—

(a) Attacks and destructive acts, including sabotage, directed against the facility by an enemy of the United States, whether a foreign government or other person; or

(b) Use or deployment of weapons incident to U.S. defense activities.

§ 53.115 Rights related to special nuclear material.

(a) No right to the SNM will be conferred by a license issued under this part except as may be defined by the license.

(b) Neither a license issued under this part, nor any right thereunder, nor any right to utilize or produce SNM may be transferred, assigned, or disposed of in any manner, either voluntarily or involuntarily, directly or indirectly, through transfer of control of the license to any person, unless the Commission, after securing full information, finds that the transfer is in accordance with the provisions of the Act and gives its consent in writing.

§ 53.117 License suspension and rights of recapture.

Any license issued under this part must be subject to suspension and to the rights of recapture of the material or control of the facility reserved to the Commission under section 108 of the Act in a state of war or national emergency declared by Congress.

§ 53.120 Information collection requirements: OMB approval.

(a) The NRC has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq. ). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0274.

(b) The approved information collection requirements contained in this part appear in §§ 53.070, 53.080, 53.240, 53.410, 53.420, 53.425, 53.430, 53.440, 53.450, 53.480, 53.500, 53.540, 53.605, 53.610, 53.620, 53.700, 53.710, 53.715, 53.720, 53.730, 53.780, 53.785, 53.805, 53.810, 53.815, 53.830, 53.850, 53.855, 53.865, 53.870, 53.875, 53.880, 53.910, 53.1010, 53.1020, 53.1030, 53.1045, 53.1060, 53.1070, 53.1075, 53.1080, 53.1100, 53.1109, 53.1115, 53.1130, 53.1140, 53.1144, 53.1146, 53.1173, 53. 1182, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210, 53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1254, 53.1257, 53,1263, 53.1270, 53.1276, 53.1279, 53.1282, 53.1288, 53.1295, 53.1300, 53.1306, 53.1309, 53.1312, 53.1327, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1384, 53.1410, 53.1413, 53.1416, 53.1419, 53.1437, 53.1449, 53.1452, 53.1458, 53.1470, 53.1505, 53.1510, 53.1515, 53.1525, 53.1530, 53.1535, 53.1540, 53.1545, 53.1550, 53.1560, 53.1565, 53.1570, 53.1575, 53.1580, 53.1620, 53.1630, 53.1645, 53.1690, 53.1720.

(c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. The information collection requirement and the control numbers under which it is approved are as follows:

(1) In §§ 53.765, 53.770, 53.780, and 53.795, NRC Form 396 is approved under control number 3150-0024.

(2) In §§ 53.775 and 53.795, NRC Form 398 is approved under control number 3150-0090.

(3) In § 53.1640, NRC Form 366 is approved under control number 3150-0104.

(4) In § 53.1630, NRC Form 361S is approved under control number 3150-0238.

(5) In § 53.1650, International Atomic Energy Agency Design Information Questionnaire forms are approved under control number 3150-0056.

(6) In § 53.1650, DOC/NRC Form AP-A and associated forms are approved under control numbers 0694-0135.

§ 53.210 Safety criteria for design-basis accidents.

Design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analyses of design-basis accidents (DBAs) in accordance with § 53.240 demonstrate the following:

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

(b) 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) TEDE.

1

1 The use of 25 rem TEDE is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.

§ 53.220 Safety criteria for licensing-basis events other than design-basis accidents.

Design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analysis of licensing-basis events (LBEs) other than DBAs in accordance with § 53.240 demonstrate the following:

(a) Plant structures, systems, and components (SSCs), personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs in accordance with §§ 53.240 and 53.450(e), and provide measures for defense in depth in accordance with § 53.250; and

(b) The analysis of risks to public health and safety resulting from LBEs other than DBAs under § 53.450(e) includes comprehensive risk metrics that satisfy associated risk performance objectives that are acceptable to the U.S. Nuclear Regulatory Commission (NRC) and provide an appropriate level of safety.

§ 53.230 Safety functions.

(a) The primary safety function is limiting the release of radioactive materials from the facility and must be maintained during normal operation and for LBEs over the life of the plant.

(b) Additional safety functions needed to support the retention of radioactive materials during LBEs—such as controlling reactivity, heat generation, heat removal, and chemical interactions—must be identified for each commercial nuclear plant.

(c) The primary and additional safety functions are required to satisfy the safety criteria defined in §§ 53.210 and 53.220 and must be fulfilled by the design features, human actions, and programmatic controls specified throughout this part.

§ 53.240 Licensing-basis events.

(a) Licensing-basis events must be identified for each commercial nuclear plant and analyzed under § 53.450 to demonstrate that the safety requirements in this subpart have been satisfied.

(b) The identified LBEs, ranging from anticipated event sequences to very unlikely event sequences, must collectively address appropriate risk-informed combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards.

(c) The analysis of LBEs must—

(1) Include analysis of one or more DBAs under § 53.450(f);

(2) Confirm the adequacy of design features and programmatic controls needed to satisfy the safety criteria defined in §§ 53.210 and 53.220, and

(3) Establish related functional requirements for plant SSCs, personnel, and programs.

§ 53.250 Defense in depth.

(a) Measures must be taken for each commercial nuclear plant to ensure appropriate defense in depth is provided to compensate for uncertainties in the analysis of the safety criteria such that there is reasonable assurance that the safety criteria in this subpart are met over the life of the plant.

(b) The uncertainties that must be addressed under paragraph (a) of this section include those related to the state of knowledge and modeling capabilities, the ability of barriers to limit the release of radioactive materials from the facility during LBEs other than DBAs, the reliability and performance of plant SSCs and personnel, and the effectiveness of programmatic controls.

(c) The safety analysis may not exclusively rely upon a single engineered design feature, human action, or programmatic control, no matter how robust, to address the range of LBEs other than DBAs.

§ 53.260 Normal operations.

Holders of licenses to operate commercial nuclear plants under this part must control public doses and dose rates in unrestricted areas from normal plant operations to meet the requirements in 10 CFR part 20.

§ 53.270 Protection of plant workers.

Holders of licenses to operate commercial nuclear plants under this part must control occupational doses to meet the requirements in 10 CFR part 20.

§ 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 §§ 53.210 and 53.220.

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

§ 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 § 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 § 53.210, and to maintain consistency with analyses required by § 53.450(f).

§ 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 § 53.510 without losing the capability to perform the safety functions identified under § 53.230. Specific requirements for earthquake engineering are included in § 53.480.

§ 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 § 53.220; and

(2) The evaluation criteria in § 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 § 53.220; and

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

§ 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 § 53.850.

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

§ 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 § 53.270 can be met.

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

§ 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 § 53.400 meets the defined functional design criteria required by §§ 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 § 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 § 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 § 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 § 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 § 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 § 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 § 20.1406 of this chapter.

(m)(1) Each commercial nuclear plant must include criticality monitoring capabilities meeting the requirements of either § 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 § 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 § 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 § 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.

§ 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 § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 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 § 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 § 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 § 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 §§ 53.220 and the analyses of licensing-basis events other than design-basis accidents (DBAs) under § 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 § 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 § 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 § 53.220 and provide defense in depth as required by § 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 § 53.240 must include analysis of DBAs that address possible challenges to the safety functions identified under § 53.230. The events selected as DBAs must be those that, if not terminated, have the potential for exceeding the safety criteria in § 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 § 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 § 53.230.

(3) The analysis must conservatively demonstrate compliance with the safety criteria in § 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 § 53.230; or

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

(2) [Reserved]

(3) Dose to members of the public. Measures taken under § 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.

§ 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 § 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 § 53.450 to demonstrate meeting the safety criteria in §§ 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 § 53.450 to demonstrate meeting the applicable criteria in §§ 53.210, 53.220, and 53.450(e).

§ 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 § 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 § 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 § 53.510(c), by taking into consideration the functional design criteria of SSCs in accordance with §§ 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 § 53.410 or § 53.420; and

(B) Structures, systems, and components classified as SR or NSRSS commensurate with safety significance in accordance with § 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 § 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 § 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 § 53.450 must address seismic hazards and related SSC responses in determining that the safety criteria defined in § 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 § 53.220 and to maintain consistency with analyses required by § 53.450 when accounting for the site-specific frequencies and magnitudes of earthquakes for a commercial nuclear plant.

§ 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 §§ 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 §§ 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 § 53.460, including, where applicable, the quality assurance requirements from appendix B of part 50 of this chapter.

§ 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 § 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 §§ 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.

§ 53.520 Site characteristics.

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

§ 53.530 Population-related considerations.

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

(a) The offsite radiological consequences estimated by the analyses required by § 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.

§ 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 §§ 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.

§ 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.

§ 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 § 53.1452(g).

(a) Definitions. The definitions in § 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 § 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 § 73.1205 of this chapter, under this section, or under § 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.

§ 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 § 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 §§ 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 § 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 § 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 § 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 § 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 § 53.620(d)(1)), a cybersecurity program that meets the requirements of § 73.54 or § 73.110 of this chapter, a physical security program that meets the requirements of § 73.55 or § 73.100 of this chapter, and an access authorization program that meets the requirements of § 73.56 or § 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 § 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 § 53.1440 or the equivalent verifications needed to support the issuance of an operating license under § 53.1387.

§ 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 §§ 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 § 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 § 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 § 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 § 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 § 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 § 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.

§ 53.700 Operational objectives.

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

(a) Each holder of an operating license (OL) or combined license (COL) under this part develops, implements, and maintains controls for plant structures, systems, and components (SSCs), responsibilities of personnel, and plant programs during the operating life of each commercial nuclear plant such that the requirements defined in subpart B are satisfied. More specifically:

(1) Under § 53.710 through § 53.730, each holder of an OL or COL under this part must maintain the capabilities, availability, and reliability of plant SSCs to ensure that the safety functions identified in § 53.230 will be performed if called upon during licensing-basis events (LBEs).

(2) Under § 53.725 through § 53.830, each holder of an OL or COL under this part must ensure that personnel have adequate knowledge and skills to perform their assigned duties that support the performance of the safety functions identified in § 53.230.

(3) Under § 53.845 through § 53.910, each holder of an OL or COL under this part must implement plant programs sufficient to ensure that the safety functions identified in § 53.230 will be performed if called upon during normal operations and LBEs.

(b) [Reserved]

§ 53.710 Maintaining capabilities and availability of structures, systems, and components.

Measures must be provided for each commercial nuclear plant licensed under this part such that the capabilities, availability, and reliability of plant SSCs, when combined with corresponding programmatic controls and human actions, provide that the safety criteria defined in §§ 53.210 and 53.220 will be met.

(a) Technical specifications must be developed, implemented, and maintained that define conditions or limitations on plant operations that are necessary to ensure that safety-related (SR) SSCs can fulfill the safety functions identified under § 53.230 and support meeting the safety criteria of § 53.210. The technical specifications must describe the following requirements:

(1) Limits on the inventory of radioactive materials within the reactor system and supporting systems with the potential, individually or collectively, to cause a release exceeding the safety criteria in § 53.210 as a result of a design-basis accident analyzed in accordance with § 53.450(f).

(2) Operating limits for the facility that if exceeded could lead to a failure to perform a required safety function necessary to demonstrate compliance with the safety criteria in § 53.210.

(3) For each SSC classified as SR in accordance with § 53.460, technical specifications must define—

(i) Limiting conditions for operation. Limiting conditions for operation are the lowest functional capability or performance levels of SR SSCs required to ensure that the design-basis accidents analyzed in accordance with § 53.450(f) satisfy the safety criteria of § 53.210. When a limiting condition for operation is not met, the licensee must shut down the plant or follow any remedial action permitted by the technical specifications until the condition can be met.

(ii) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that the limiting conditions for operation will be met.

(4) Design elements to be included are those elements of the plant such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (a)(1) through (3) of this section.

(5) Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the plant in a safe manner. Each licensee must submit any reports to the Commission pursuant to approved technical specifications under § 53.040.

(b) Control measures on plant operations, including availability controls, must be developed and implemented to ensure that the configurations and special treatments for SR SSCs and non-safety-related but safety-significant (NSRSS) SSCs provide the capabilities, availability, and reliability required to demonstrate compliance with the criteria of §§ 53.220 and 53.450(e).

1

The control measures must—

1 The comprehensive risk metrics and related risk performance objectives established under § 53.220 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.

(1)(i) Identify who within the licensee's organization has authority to make configuration changes;

(ii) Establish processes to make configuration changes to NSRSS SSCs; and

(iii) Establish processes to ensure that all organizations of the commercial nuclear plant affected by the configuration changes are formally notified and approve of the change.

(2) Describe how the special treatments for each NSRSS SSC and special treatments for SR SSCs beyond those under paragraph (a) of this section will be established and maintained over the operating life of the commercial nuclear plant.

§ 53.715 Maintenance, repair, and inspection programs.

(a) A program to control maintenance activities and monitor the performance or condition of SR and NSRSS SSCs must be developed, implemented, and maintained.

(b) Whenever a licensee determines through activities related to maintenance, repair, and inspection of SSCs, the activities under § 53.710, or otherwise that the performance or condition of an SR or NSRSS SSC does not demonstrate compliance with established special treatments or performance goals related to capabilities, availability, or reliability, the licensee must take appropriate corrective action.

(c) Performance and condition monitoring activities and associated goals and preventive maintenance activities must be evaluated at least every 24 months. The evaluations must take into account, where practical, industry-wide operating experience. Adjustments must be made where necessary to ensure that the objective of preventing failures of SSCs through maintenance is appropriately balanced against the objective of minimizing unavailability of SSCs due to monitoring or preventive maintenance.

(d) Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee must assess and manage the increase in risk that may result from the proposed maintenance activities.

§ 53.720 Response to seismic events.

If vibratory ground motion exceeding that of the operating basis earthquake Ground Motion or significant plant damage due to vibratory ground motion occurs, the licensee must shut down the commercial nuclear plant. If structures, systems, or components necessary for the safe shutdown of the commercial nuclear plant are not available after the occurrence of this vibratory ground motion, the licensee must consult with the Commission and must propose a plan for the timely, safe shutdown of the commercial nuclear plant. Prior to resuming operations, the licensee must demonstrate to the Commission that those features necessary for continued operation without undue risk to the health and safety of the public or necessary to maintain the licensing basis of the commercial nuclear plant were either not functionally damaged or have been repaired.

§ 53.725 General staffing, training, personnel qualifications, and human factors requirements.

(a) Two classes of commercial nuclear plants. Commercial nuclear plants licensed under this part are either of the class of self-reliant-mitigation facilities or of interaction-dependent-mitigation facilities, based upon the similarity of operating and technical characteristics of the plants in the class. A commercial nuclear plant is a self-reliant-mitigation facility if the U.S. Nuclear Regulatory Commission (NRC) determined as part of its approval of the OL or COL for that plant that its design demonstrates compliance with the criteria of § 53.800(a)(1) through (a)(5). Otherwise, the commercial nuclear plant is an interaction-dependent-mitigation facility.

(b) Purpose and applicability. The regulations in §§ 53.725 through 53.830 address areas related to staffing, training, personnel qualifications, and human factors engineering for applicants for or holders of OLs or COLs under this part. These regulations are organized as follows:

(1) Sections 53.725 through 53.745 address general requirements for staffing, training, personnel qualifications, and human factors engineering. The regulations within these sections are applicable to all applicants for or holders of OLs or COLs under this part, except where specifically stated otherwise.

(2) Sections 53.760 through 53.795 address operator and senior operator licensing requirements. The regulations within these sections are applicable to those applicants for or holders of OLs or COLs under this part for interaction-dependent-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070.

(3) Sections 53.800 through 53.820 address generally licensed reactor operator requirements. The regulations within these sections are in lieu of §§ 53.760 through 53.795 for those applicants for or holders of OLs or COLs under this part for self-reliant-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070.

(4) Section 53.830 provides general personnel training requirements. The regulations within this section are applicable to all applicants for or holders of OLs or COLs under this part.

(c) Definitions. When used in §§ 53.725 through 53.830, applicant refers to an applicant for an operator or senior operator license; licensee refers to the holder of an operator, senior operator, or generally licensed reactor operator license; and facility licensee refers to the licensee for the commercial nuclear plant where the applicant would be licensed or the licensee is licensed. As also used in §§ 53.725 through 53.830—

Automation means a device or system that accomplishes (partially or fully) a function or task.

Auxiliary operator means any individual who operates components of a commercial nuclear plant but does not manipulate controls or direct the manipulation of controls of the plant and is not required to be licensed under the provisions of this part.

Controls when used with respect to a nuclear reactor means apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor.

Generally licensed reactor operator means any individual licensed under the provisions of § 53.810 to manipulate controls of a self-reliant-mitigation facility and to direct the licensed activities of generally licensed reactor operators.

Interaction-dependent-mitigation facility means a commercial nuclear plant design other than one that demonstrates compliance with the operating and technical characteristics defined under § 53.800.

Load following means a commercial nuclear plant automatically changing its output to match expected demand in response to externally originated instructions or signals.

Operator means any individual licensed under the provisions of §§ 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility.

Performance testing means testing conducted to verify a simulation facility's performance as compared to actual or predicted reference plant performance.

Reference plant means the specific commercial nuclear plant, or plant design for facilities which are not yet constructed, on which a simulation facility's configuration, system control arrangement, and design data are based.

Self-reliant-mitigation facility means a commercial nuclear plant design that demonstrates compliance with the operating and technical characteristics defined under § 53.800.

Senior operator means any individual licensed under the provisions of §§ 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility and to direct the licensed activities of operators.

Simulation facility means an interface designed to provide a realistic imitation of the operation of a commercial nuclear plant used for the administration of examinations, for training, and/or to demonstrate compliance with experience requirements for applicants or licensees. A simulation facility may rely, in whole or part, upon the physical utilization of the reference plant itself.

Systems approach to training means a training program that includes the following five elements:

(i) Systematic analysis of the jobs to be performed.

(ii) Learning objectives derived from the analysis which describe desired performance after training.

(iii) Training design and implementation based on the learning objectives.

(iv) Evaluation of trainee mastery of the objectives during training.

(v) Evaluation and revision of the training based on the performance of trained personnel in the job setting.

§ 53.726 Communications.

(a) An applicant or licensee or facility licensee must submit any communication or report required by the regulations contained within §§ 53.725 through 53.830 and must submit any application filed under these regulations to the Commission.

(b) Each facility licensee that is required to comply with the requirements of §§ 53.760 through 53.795 ( i.e., interaction-dependent-mitigation facilities) must notify the appropriate NRC contact within 30 days of the following in regard to a licensed operator or senior operator:

(1) Permanent reassignment from the position for which the facility licensee has certified the need for a licensed operator or senior operator under § 53.775(a)(1);

(2) Termination of any operator or senior operator; or

(3) Permanent disability or illness as required under § 53.770.

§ 53.728 Completeness and accuracy of information.

Information provided to the Commission by an applicant for an operator or senior operator license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee must be complete and accurate in all material respects.

§ 53.730 Defining, fulfilling, and maintaining the role of personnel in ensuring safe operations.

Each applicant for or holder of an OL or COL for a commercial nuclear plant under this part must comply with the following:

(a) Human factors engineering design requirements. The plant design must reflect state-of-the-art human factors engineering 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.

(b) Human system interface design requirements. The plant design must provide for the following to support operating personnel in monitoring plant conditions and responding to plant events:

(1) Features for displaying to operating personnel a minimum set of parameters that define the safety status of the plant and are capable of displaying both the full range of important plant parameters and data trends on demand, as well as indicating when process limits are being approached or exceeded;

(2) Automatic indication of the bypassed and operable status of safety systems;

(3) Direct indication of SSC status that relates to the ability of the SSC to perform its safety function, such as relief and safety valve position ( i.e., open or closed) for barriers important to fulfilling safety functions with such devices, and ultimate heat sink and cooling system status and availability;

(4) Instrumentation to measure, record, and display key plant parameters related to the performance of SSCs and the integrity of barriers important to fulfilling safety functions to support operators in monitoring plant conditions and responding to plant events. Examples include temperatures and pressures within important systems or structures, core or fuel system conditions (including possible damage states), temperatures and levels associated with cooling functions, combustible gas concentrations, radiation levels in systems and within structures, and radioactive effluent releases;

(5) Leakage control and detection in the design of systems that pass through barriers important to fulfilling safety functions for the release of radionuclides. An example is an SSC that penetrates a containment structure that might contain radioactive materials that could contribute to the source term during an accident;

(6) Monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of normal operating and accident conditions; and

(7) For self-reliant-mitigation facilities, the plant design must also provide the generally licensed reactor operators with the capability to do the following:

(i) Receive plant operating data, including reactor parameters and information needed for the evaluation of emergency conditions.

(ii) Promptly dispatch operations and maintenance personnel.

(iii) Immediately implement responsibilities under the facility emergency plan, as applicable.

(8) For both interaction-dependent and self-reliant mitigation facilities, the plant design must provide licensed operators with the capability of immediately initiating a reactor shutdown from their location.

(c) Concept of operations. A concept of operations that is of sufficient scope and detail to address the following must be provided:

(1) Plant goals;

(2) The roles and responsibilities of operating personnel and automation (or any combination thereof) that are responsible for completing plant functions;

(3) Staffing, qualifications, and training;

(4) The management of normal operations;

(5) The management of off-normal conditions and emergencies;

(6) The management of maintenance and modifications; and

(7) The management of tests, inspections, and surveillances.

(d) Functional requirements analysis and function allocation. A functional requirements analysis and a function allocation must be provided that are sufficient to demonstrate compliance with the following:

(1) The functional requirements analysis must address how safety functions and functional safety criteria are satisfied; and

(2) The function allocation must describe how the safety functions will be assigned to human action, automation, active safety features, passive safety features, and/or inherent safety characteristics.

(e) Operating experience. A program, during construction and during operation, as applicable, for evaluating and applying operating experience must be developed, implemented, and maintained.

(f) Staffing plan. A staffing plan must be developed and comply with the following:

(1) The staffing plan must include a description of how engineering expertise will be available to the on-shift operating personnel during all plant conditions, to assist if they encounter a situation not covered by procedures or training. Engineering expertise includes familiarity with the operation of the plant for which the expertise is provided and one of the following:

(i) A bachelor's degree in engineering, engineering technology, or physical science from an institution accredited by a U.S. Government recognized accrediting body or equivalent; or

(ii) A Professional Engineer's license from a U.S. State or territory.

(2) Applicants for or holders of OLs or COLs for interaction-dependent-mitigation facilities must include within their staffing plans a description of how the proposed numbers, positions, and qualifications of operators and senior operators across all modes of plant operations will be sufficient to ensure that plant safety functions will be maintained. This description must be supported by human factors engineering analyses and assessments.

(3) Applicants for or holders of OLs or COLs for self-reliant-mitigation facilities must include within their staffing plans a description of how generally licensed reactor operator staffing that is both sufficient to continually monitor the operations of fueled reactors and to provide for a continuity of responsibility for facility operations at all times during the operating phase will be maintained.

(4) Applicants for or holders of OLs or COLs under this part must include within their staffing plans a description of how the positions and responsibilities of personnel contained within those plans will adequately satisfy necessary support functions within areas such as plant operations, equipment surveillance and maintenance, radiological protection, chemistry control, fire brigades, engineering, security, and emergency response.

(5) The staffing plan must be approved by the NRC as part of its approval of the OL or COL for the plant. The approved staffing plan is subject to the requirements of § 53.1565.

(g) Training, examination, and proficiency programs. Develop, implement, and maintain programs that comply with the following requirements. These programs must be approved by the NRC as part of its approval of the OL or COL for the plant:

(1) For those applicants for or holders of OLs or COLs for interaction-dependent-mitigation facilities:

(i) The operator licensing initial training program required under § 53.780(a);

(ii) The operator licensing initial examination program required under § 53.780(b);

(iii) The operator licensing requalification program required under § 53.780(c); and

(iv) The operator proficiency program required under § 53.780(g).

(2) For those applicants for or holders of OLs or COLs for self-reliant-mitigation facilities, the generally licensed reactor operator training, examination, and proficiency programs required under § 53.815.

(3) The operator licensing requalification programs required under § 53.780(c) or § 53.815(b) must be implemented upon commencing the administration of initial examinations under the operator licensing examination program required under § 53.780(b) or § 53.815(b), respectively.

§ 53.735 General exemptions.

The regulations in §§ 53.725 through 53.830 do not require a license for an individual who—

(a) Under the direction and in the presence of an operator or senior operator or a generally licensed reactor operator, as appropriate, manipulates the controls of a commercial nuclear plant as a part of the individual's training in a facility licensee's training program as approved by the Commission to qualify for an operator or senior operator license or a generally licensed reactor operator license there, as appropriate, under these regulations; or

(b) Under the direction and in the presence of a senior operator or generally licensed reactor operator, as appropriate, manipulates the controls of a commercial nuclear plant to load or unload the fuel into, out of, or within the reactor vessel while the reactor is not operating.

§ 53.740 Facility licensee requirements—general.

(a) Facility licensees must demonstrate compliance with the requirements of either §§ 53.760 through 53.795 for interaction-dependent-mitigation facilities or §§ 53.800 through 53.820 for self-reliant-mitigation facilities.

(b) The facility licensee must maintain the staffing complement described under its approved facility staffing plan until such time as the permanent cessation of operations and permanent removal of fuel from the reactor vessel has been certified as described under § 53.1070. The approved staffing plan is subject to the requirements of § 53.1565.

(c) Except as provided under § 53.735, the facility licensee may not permit the manipulation of the controls of a commercial nuclear plant by anyone who is not an operator or senior operator or generally licensed reactor operator, as appropriate.

(d) Facility licensees for interaction-dependent-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070 must designate senior operators to be responsible for supervising the licensed activities of operators.

(e) Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor, must be manipulated only while plant conditions are being monitored by an individual who is an operator or senior operator or a generally licensed reactor operator, as appropriate.

(f)(1) Load following is permitted if at least one of the following is immediately capable of refusing demands when they could challenge the safe operation of the plant or when precluded by the plant equipment conditions:

(i) The actuation of an automatic protection system that utilizes setpoints more conservative than those otherwise credited for the purposes of reactor protection; or

(ii) An automated control system; or

(iii) An operator or senior operator or a generally licensed reactor operator, as appropriate.

(2) The provisions of paragraph (e) of this section do not apply during load following operations.

(g)(1) Facility licensees for interaction-dependent-mitigation facilities must have present during alteration of the core (including fuel loading or transfer) an individual holding a senior operator license, or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the facility licensee must not assign other duties to this person.

(2) Facility licensees for self-reliant-mitigation facilities must have present during alteration of the core (including fuel loading or transfer) an individual holding a generally licensed reactor operator license to directly supervise the activity and, during this time, the facility licensee must not assign other duties to this person.

(3) The provisions of paragraphs (g)(1) and (2) of this section do not apply to core alterations performed as part of refueling operations while a facility that is capable of online refueling is operating at power.

(h) Facility licensees may take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent. Such facility licensee action must be approved, as a minimum, by a senior operator or a generally licensed reactor operator, as applicable, or, after certifying the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070 by a certified fuel handler, senior operator, or generally licensed reactor operator, as applicable, prior to taking the action.