{"database": "openregs", "table": "cfr_sections", "is_view": false, "human_description_en": "where part_number = 53 and title_number = 10 sorted by section_id", "rows": [["10:10:2.0.1.1.3.0.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "", "", "", "\u00a7 53.000 Purpose.", "NRC", "", "", "", "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 \u00a7 53.050."], ["10:10:2.0.1.1.3.1.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.015 Scope.", "NRC", "", "", "", "Subpart A provides general provisions applicable to all applicants and licensees subject to the rules of this part."], ["10:10:2.0.1.1.3.1.62.10", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.100 Jurisdictional limits.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.1.62.11", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.110 Attacks and destructive acts.", "NRC", "", "", "", "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\u2014\n\n(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\n\n(b) Use or deployment of weapons incident to U.S. defense activities."], ["10:10:2.0.1.1.3.1.62.12", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.115 Rights related to special nuclear material.", "NRC", "", "", "", "(a) No right to the SNM will be conferred by a license issued under this part except as may be defined by the license.\n\n(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."], ["10:10:2.0.1.1.3.1.62.13", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.117 License suspension and rights of recapture.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.1.62.14", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.120 Information collection requirements: OMB approval.", "NRC", "", "", "", "(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.\n\n(b) The approved information collection requirements contained in this part appear in \u00a7\u00a7 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.\n\n(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:\n\n(1) In \u00a7\u00a7 53.765, 53.770, 53.780, and 53.795, NRC Form 396 is approved under control number 3150-0024.\n\n(2) In \u00a7\u00a7 53.775 and 53.795, NRC Form 398 is approved under control number 3150-0090.\n\n(3) In \u00a7 53.1640, NRC Form 366 is approved under control number 3150-0104.\n\n(4) In \u00a7 53.1630, NRC Form 361S is approved under control number 3150-0238.\n\n(5) In \u00a7 53.1650, International Atomic Energy Agency Design Information Questionnaire forms are approved under control number 3150-0056.\n\n(6) In \u00a7 53.1650, DOC/NRC Form AP-A and associated forms are approved under control numbers 0694-0135."], ["10:10:2.0.1.1.3.1.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.020 Definitions.", "NRC", "", "", "[91 FR 15794, Mar. 30, 2026, as amended at 91 FR 18773, Apr. 13, 2026]", "As used in this part:\n\nAnticipated 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.\n\nApplicant  means a person applying for a license, permit, or other form of Commission permission or approval under this part.\n\nCertified fuel handler  means, for a commercial nuclear plant, either\u2014\n\n(1) A non-licensed operator who has qualified in accordance with a fuel handler training program approved by the Commission; or\n\n(2) A non-licensed operator who demonstrates compliance with the following criteria:\n\n(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 \u00a7 53.830, and\n\n(ii) Is responsible for decisions on\u2014\n\n(A) Safe conduct of decommissioning activities,\n\n(B) Safe handling and storage of spent fuel; and\n\n(C) Appropriate response to plant emergencies.\n\nCombined license (COL)  means a combined construction permit (CP) and operating license (OL) with conditions for a commercial nuclear plant issued under this part.\n\nCommercial 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.\n\nCommercial 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 \u00a7 50.2 of this chapter.\n\nCommission  means the U.S. Nuclear Regulatory Commission (NRC) or its duly authorized representatives.\n\nConstruction  means the activities in paragraph (1) of this definition and does not mean the activities in paragraph (2) of this definition.\n\n(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\u2014\n\n(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;\n\n(ii) SSCs necessary to comply with 10 CFR part 73; or\n\n(iii) Onsite emergency facilities necessary to comply with \u00a7 53.855.\n\n(2) Construction does not include\u2014\n\n(i) Changes for temporary use of the land for public recreational purposes;\n\n(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;\n\n(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;\n\n(iv) Erection of fences and other access control measures;\n\n(v) Excavation;\n\n(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;\n\n(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);\n\n(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\n\n(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.\n\nCustom combined license (custom COL)  means a COL that does not reference a standard design approval, standard design certification, or manufacturing license.\n\nDecommission or decommissioning  means to remove a plant or site safely from service and reduce residual radioactivity to a level that permits\u2014\n\n(1) Release of the property for unrestricted use and termination of the license; or\n\n(2) Release of the property under restricted conditions and termination of the license.\n\nDefense 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.\n\nDesign-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.\n\nDesign-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.\n\nDesign 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.\n\nEarly 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.\n\nElectric 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 \u201celectric utility.\u201d\n\nEvent 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.\n\nExclusion 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.\n\nFission 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.\n\nFuel  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.\n\nFunctional 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 \u00a7 53.210. For NSRSS SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in \u00a7 53.220.\n\nLicense,  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.\n\nLicensee  means a person who is authorized to conduct activities under a license issued under this part by the Commission.\n\nLicensing-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.\n\nLicensing-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.\n\nLow-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.\n\nMajor 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 \u00a7 61.55 of this chapter.\n\nMajor feature of the emergency plans  means an aspect of those plans necessary to:\n\n(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\n\n(2) Describe the emergency planning zones as required in 10 CFR 53.1109(g).\n\nManufactured 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.\n\nManufacturing license  means a license issued under this part that authorizes the manufacture of manufactured reactors but not its construction, installation, or operation.\n\nNon-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.\n\nNon-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.\n\nPerson  means\u2014\n\n(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\n\n(2) Any legal successor, representative, agent, or agency of the foregoing.\n\nPopulation center distance  means the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents.\n\nProgrammatic 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 \u00a7 53.845 and are specified in an application for a requested activity of the Commission.\n\nQuality 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.\n\nSafety criteria  means performance-based metrics that establish a level of safety provided in requirements in \u00a7\u00a7 53.210 and 53.220.\n\nSafety-related structures, systems, or components  means those SSCs that are relied upon to demonstrate compliance with the safety criteria in \u00a7 53.210 and warrant special treatment.\n\nSmall modular reactor  means a power reactor, which may be of modular design as defined in \u00a7 52.1 of this chapter, licensed under this part to produce heat energy up to 1,000 megawatts thermal per module.\n\nSite 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.\n\nSite 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.\n\nSource 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.\n\nSpecial nuclear material (SNM)  means:\n\n(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\n\n(2) Any material artificially enriched by any of the foregoing, but does not include source material.\n\nSpecial 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 \u00a7 53.450 to demonstrate compliance with the safety criteria in \u00a7\u00a7 53.210 for SR SSCs and 53.220 for SR and NSRSS SSCs.\n\nStandard 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.\n\nStandard 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.\n\nStandard 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.\n\nTotal effective dose equivalent  means the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).\n\nUtilization facility  means any commercial nuclear reactor other than one designed or used primarily for the formation of plutonium or uranium-233.\n\nUnlikely 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.\n\nVery 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."], ["10:10:2.0.1.1.3.1.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.030 [Reserved]", "NRC", "", "", "", ""], ["10:10:2.0.1.1.3.1.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.040 Written communications.", "NRC", "", "", "", "(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  MSHD.Resource@nrc.gov;  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.\n\n(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).\n\n(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.\n\n(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.\n\n(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.\n\n(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:\n\n(i) Physical security plan;\n\n(ii) Safeguards contingency plan;\n\n(iii) Cybersecurity plan;\n\n(iv) Change to security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan made without prior Commission approval under \u00a7 53.1565; and\n\n(v) Application for amendment of physical security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan under \u00a7 53.1510.\n\n(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:\n\n(i) Emergency plan;\n\n(ii) Change to an emergency plan under \u00a7 53.1565; and\n\n(iii) Emergency implementing procedures under \u00a7 53.855.\n\n(6)  Updated Final Safety Analysis Report.  An updated Final Safety Analysis Report or replacement pages under \u00a7 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.\n\n(7)  Quality assurance related submissions.  (i) A change to the Safety Analysis Report QA program description under \u00a7 53.1565, or a change to a licensee's NRC-accepted QA topical report under \u00a7 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.\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.1.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.050 Deliberate misconduct.", "NRC", "", "", "", "(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\u2014\n\n(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\n\n(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.\n\n(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.\n\n(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\u2014\n\n(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\n\n(2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor."], ["10:10:2.0.1.1.3.1.62.6", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.060 Employee protection.", "NRC", "", "", "", "(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.\n\n(1) The protected activities include but are not limited to\u2014\n\n(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;\n\n(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;\n\n(iii) Requesting the NRC to institute action against his or her employer for the administration or enforcement of these requirements;\n\n(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\n\n(v) Assisting or participating in, or being about to assist or participate in, these activities.\n\n(2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.\n\n(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.\n\n(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.\n\n(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\u2014\n\n(1) Denial, revocation, or suspension of the license or standard design approval;\n\n(2) Withdrawal or revocation of a proposed or final standard design certification;\n\n(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\n\n(4) Other enforcement action.\n\n(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.\n\n(e)(1) Each holder or applicant for a license or design approval, must prominently post the revision of NRC Form 3, \u201cNotice to Employees,\u201d referenced in \u00a7 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.\n\n(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  Forms.Resource@nrc.gov,  or by visiting the NRC's online library at  https://www.nrc.gov/reading-rm/doc-collections/forms/.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.1.62.7", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.070 Completeness and accuracy of information.", "NRC", "", "", "", "(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.\n\n(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."], ["10:10:2.0.1.1.3.1.62.8", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.080 Specific exemptions.", "NRC", "", "", "", "(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.\n\n(b) The Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever\u2014\n\n(1) Application of the regulation in the particular circumstances conflicts with other rules or requirements of the Commission;\n\n(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;\n\n(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;\n\n(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;\n\n(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\n\n(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.\n\n(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:\n\n(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;\n\n(2) Whether redress of any adverse environment impact from conduct of the proposed activities can reasonably be effective should such redress be necessary;\n\n(3) Whether conduct of the proposed activities would foreclose subsequent adoption of alternatives; and\n\n(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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.1.62.9", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "A", "Subpart A\u2014General Provisions", "", "\u00a7 53.090 Standards for review.", "NRC", "", "", "", "(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:\n\n(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.\n\n(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.\n\n(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.\n\n(4) Any applicable requirements of 10 CFR part 51 have been satisfied.\n\n(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.\n\n(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.\n\n(1) The NRC will\u2014\n\n(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;\n\n(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\n\n(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 .\n\n(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.\n\n(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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.10.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1600 General information.", "NRC", "", "", "", "Each applicant and licensee under this part must ensure that U.S. Nuclear Regulatory Commission (NRC) inspectors have unfettered access to sites and facilities licensed or proposed to be licensed in \u00a7 53.1610, must maintain records and make reports to the NRC in accordance with requirements in \u00a7\u00a7 53.1620 through 53.1650, must satisfy financial qualification and reporting requirements in \u00a7\u00a7 53.1660 through 53.1700, and must obtain and maintain required financial protections in case of an accident in \u00a7\u00a7 53.1720 and 53.1730."], ["10:10:2.0.1.1.3.10.62.10", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1680 [Reserved]", "NRC", "", "", "", ""], ["10:10:2.0.1.1.3.10.62.11", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1690 Licensee's change of status; financial qualifications.", "NRC", "", "", "", "(a) An electric utility licensee holding an OL or COL (including a renewed license) for a commercial nuclear plant, no later than seventy-five (75) days prior to ceasing to be an electric utility in any manner not involving a license transfer under \u00a7 53.1399 or \u00a7 53.1456 must provide the NRC with the financial qualifications information that would be required for obtaining an initial OL under this part. The financial qualifications information must address the first full 5 years of operation after the date the licensee ceases to be an electric utility.\n\n(b)(1) Any holder of a license issued under this part must notify the appropriate NRC Regional Administrator, in writing, immediately following the filing of a voluntary or involuntary petition for bankruptcy under any chapter of title 11 (Bankruptcy) of the United States Code by or against\u2014\n\n(i) The licensee;\n\n(ii) An entity (as 11 U.S.C. 101(14) defines that term) controlling the licensee or listing the license or licensee as property of the estate; or\n\n(iii) An affiliate (as 11 U.S.C. 101(2) defines that term) of the licensee.\n\n(2) This notification must indicate\u2014\n\n(i) The bankruptcy court in which the petition for bankruptcy was filed; and\n\n(ii) The date of the filing of the petition."], ["10:10:2.0.1.1.3.10.62.12", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1700 Creditor regulations.", "NRC", "", "", "", "(a) Pursuant to section 184 of the Act, the Commission consents, without individual application, to the creation of any mortgage, pledge, or other lien upon any facility not owned by the United States which is the subject of a license or upon any leasehold or other interest in such facility; provided\u2014\n\n(1) That the rights of any creditor so secured may be exercised only in compliance with and subject to the same requirements and restrictions as would apply to the licensee pursuant to the provisions of the license, the Act, and regulations issued by the Commission under the Act; and\n\n(2) That no creditor so secured may take possession of the facility pursuant to the provisions of this section prior to either the issuance of a license from the Commission authorizing such possession or the transfer of the license.\n\n(b) Any creditor so secured may apply for transfer of the license covering such facility by filing an application for transfer of the license under \u00a7 53.1570. The Commission will act upon such application under subpart I of this part.\n\n(c) Nothing contained in this regulation shall be deemed to affect the means of acquiring, or the priority of, any tax lien or other lien provided by law.\n\n(d) As used in this section\u2014\n\n(1)  License  includes any license under this part, which may be issued by the Commission with regard to a facility.\n\n(2)  Creditor  includes, without implied limitation, the trustee under any mortgage, pledge or lien on a facility made to secure any creditor, any trustee or receiver of the facility appointed by a court of competent jurisdiction in any action brought for the benefit of any creditor secured by such mortgage, pledge or lien, any purchaser of such facility at the sale thereof upon foreclosure of such mortgage, pledge, or lien or upon exercise of any power of sale contained therein, or any assignee of any such purchaser.\n\n(3)  Facility  includes, but is not limited to, a site which is the subject of an early site permit under this part, and a reactor manufactured under an ML under this part."], ["10:10:2.0.1.1.3.10.62.13", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1710 Financial protection.", "NRC", "", "", "", "Sections 53.1720 and 53.1730 set out the requirements and procedures related to licensees obtaining and maintaining insurance to cover stabilization and decontamination activities in the event of an accident and financial protection in accordance with part 140, \u201cFinancial Protection Requirements and Indemnity Agreements,\u201d of this chapter."], ["10:10:2.0.1.1.3.10.62.14", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1720 Insurance required to stabilize and decontaminate plant following an accident.", "NRC", "", "", "", "Each commercial nuclear plant licensee under this part must take reasonable steps to obtain insurance available at reasonable costs and on reasonable terms from private sources or to demonstrate that it possesses an equivalent amount of protection covering the licensee's obligation, in the event of an accident at the licensee's commercial nuclear reactor, to stabilize and decontaminate the plant and the plant site at which such an accident may occur, provided that\u2014\n\n(a) The insurance required by this section must have a minimum coverage limit for each commercial nuclear plant site of $1.06 billion, an amount based on plant-specific estimates of costs to stabilize and decontaminate a plant, or whatever amount of insurance is generally available from private sources, whichever is less. The required insurance must clearly state that, as and to the extent provided in paragraph (d)(1) of this section, any proceeds must be payable first for stabilization of the plant and next for decontamination of the plant and the plant site. If a licensee's coverage falls below the required minimum, the licensee must within 60 days take all reasonable steps to restore its coverage to the required minimum. The required insurance may, at the option of the licensee, be included within policies that also provide coverage for other risks, including, but not limited to, the risk of direct physical damage.\n\n(b)(1) With respect to policies issued or annually renewed, the proceeds of such required insurance must be dedicated, as and to the extent provided in this paragraph (b), to reimbursement or payment on behalf of the insured of reasonable expenses incurred or estimated to be incurred by the licensee in taking action to fulfill the licensee's obligation, in the event of an accident at the licensee's plant, to ensure that the plant is in, or is returned to, and maintained in, a safe and stable condition and that radioactive contamination is removed or controlled such that personnel exposures are consistent with the occupational exposure limits in 10 CFR part 20. These actions must be consistent with any other obligation the licensee may have under this chapter and must be subject to paragraph (d) of this section. As used in this section, an \u201caccident\u201d means an event that involves the release of radioactive material from its intended place of confinement within the commercial nuclear plant such that there is a present danger of release off site in amounts that would pose a threat to the public health and safety.\n\n(2) The stabilization and decontamination requirements set forth in paragraph (d) of this section must apply uniformly to all insurance policies required under this section.\n\n(c) The licensee shall report to the NRC on April 1 of each year the current levels of this insurance or financial security it maintains and the sources of this insurance or financial security.\n\n(d)(1) In the event of an accident at the licensee's plant, whenever the estimated costs of stabilizing the licensed plant and of decontaminating the plant and the plant site exceed one tenth of the minimum insurance under paragraph (a) of this section, the proceeds of the insurance required by this section must be dedicated to and used, first, to ensure that the licensed plant is in, or is returned to, and can be maintained in, a safe and stable condition so as to prevent any significant risk to the public health and safety and, second, to decontaminate the plant and the plant site in accordance with the licensee's cleanup plan as approved by order of the Director, Office of Nuclear Reactor Regulation. This priority on insurance proceeds must remain in effect for 60 days or, upon order of the Director, for such longer periods, in increments not to exceed 60 days except as provided for activities under the cleanup plan required in paragraphs (d)(3) and (d)(4) of this section, as the Director may find necessary to protect the public health and safety. Actions needed to bring the plant to and maintain the plant in a safe and stable condition may include one or more of the following, as appropriate:\n\n(i) Shutdown of the reactor(s) and other processes at the plant;\n\n(ii) Establishment and maintenance of long-term cooling with stable decay heat removal;\n\n(iii) Maintenance of sub-criticality;\n\n(iv) Control of radioactive releases; and\n\n(v) Securing of structures, systems, or components to minimize radiation exposure to onsite personnel or to the offsite public or to facilitate later decontamination or both.\n\n(2) The licensee must inform the Director, Office of Nuclear Reactor Regulation in writing when the plant is and can be maintained in a safe and stable condition so as to prevent any significant risk to the public health and safety. Within 30 days after the licensee informs the Director that the plant is in this condition, or at such earlier time as the licensee may elect or the Director may for good cause direct, the licensee must prepare and submit a cleanup plan for the Director's approval. The cleanup plan must identify and contain an estimate of the cost of each cleanup operation that will be required to decontaminate the reactor sufficiently to permit the licensee either to resume operation of the reactor or to apply to the Commission under subpart G of this part for authority to decommission the reactor and to surrender the license voluntarily. Cleanup operations may include one or more of the following, as appropriate:\n\n(i) Processing any contaminated materials generated by the accident and by decontamination operations to remove radioactive materials;\n\n(ii) Decontamination of surfaces inside the plant buildings to levels consistent with the Commission's occupational exposure limits in 10 CFR part 20, and decontamination or disposal of equipment;\n\n(iii) Decontamination or removal and disposal of internal parts, damaged fuel from the reactor coolant or fuel systems, or related process or waste systems; and\n\n(iv) Cleanup of the reactor coolant or fuel systems or related process or waste systems.\n\n(3) Following review of the licensee's cleanup plan, the Director will order the licensee to complete all operations that the Director finds are necessary to decontaminate the reactor sufficiently to permit the licensee either to resume operation of the reactor or to apply to the Commission under subpart G of this part for authority to decommission the reactor and to surrender the license voluntarily. The Director must approve or disapprove, in whole or in part for stated reasons, the licensee's estimate of cleanup costs for such operations. Such order may not be effective for more than one year, at which time it may be renewed. Each subsequent renewal order, if imposed, may be effective for not more than 6 months.\n\n(4) Of the balance of the proceeds of the required insurance not already expended to place the plant in a safe and stable condition under paragraph (b)(1) of this section, an amount sufficient to cover the expenses of completion of those decontamination operations that are the subject of the Director's order must be dedicated to such use, provided that, upon certification to the Director of the amounts expended previously and from time to time for stabilization and decontamination and upon further certification to the Director as to the sufficiency of the dedicated amount remaining, policies of insurance may provide for payment to the licensee or other loss payees of amounts not so dedicated, and the licensee may proceed to use in parallel (and not in preference thereto) any insurance proceeds not so dedicated for other purposes."], ["10:10:2.0.1.1.3.10.62.15", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1730 Financial protection requirements.", "NRC", "", "", "", "Commercial nuclear plant licensees must satisfy the applicable provisions of part 140, \u201cFinancial Protection Requirements and Indemnity Agreements,\u201d of this chapter."], ["10:10:2.0.1.1.3.10.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1610 Unfettered access for inspections.", "NRC", "", "", "", "(a) Each applicant for or holder of a manufacturing license (ML), operating license (OL), combined license (COL), construction permit (CP), or early site permit must permit inspection, by duly authorized representatives of the Commission, of its records, premises, activities, and of licensed materials in possession or use, related to the license or CP or early site permit as may be necessary to effectuate the purposes of the Atomic Energy Act of 1956, as amended, (the Act) and the Energy Reorganization Act of 1974, as amended.\n\n(b)(1) Each holder of an ML, OL, COL, or CP must, upon request by the Director, Office of Nuclear Reactor Regulation, provide rent-free office space for the exclusive use of the Commission inspection personnel. Heat, air conditioning, light, electrical outlets, and janitorial services must be furnished by each licensee and each holder of a CP. The office must be convenient to and have full access to the facility and must provide the inspectors both visual and acoustic privacy.\n\n(2) For a site or facility with an assigned resident inspector, the space provided must be adequate to accommodate a full-time inspector, a part-time secretary, and transient NRC personnel and must be generally commensurate with other office facilities at the site. For sites or facilities assigned multiple resident inspectors, additional space may be requested. The office space that is provided must be subject to the approval of the Director, Office of Nuclear Reactor Regulation. All furniture, supplies, and communication equipment will be furnished by the Commission.\n\n(3) For a site or facility without an assigned resident inspector, temporary space to accommodate periodic or special inspections must be provided. The office space must be generally commensurate with other office accommodations at the site.\n\n(4) The licensee or permit holder must afford any NRC resident inspector assigned to that site, or other NRC inspectors identified by the Regional Administrator as likely to inspect the facility, immediate unfettered access, equivalent to access provided regular plant employees, following proper identification and compliance with applicable access control measures for security, radiological protection, and personal safety.\n\n(5) The licensee or permit holder must ensure that the arrival and presence of an NRC inspector, who has been properly authorized facility access as described in paragraph (b)(4) of this section, is not announced or otherwise communicated by its employees or contractors to other persons at the facility unless specifically requested by the NRC inspector."], ["10:10:2.0.1.1.3.10.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1620 Maintenance of records, making of reports.", "NRC", "", "", "", "(a) Each holder of an ML, OL, COL, CP, or early site permit must maintain all records and make all reports, in connection with the activity, as may be required by the conditions of the license or permit or by the regulations and orders of the Commission in effectuating the purposes of the Act and the Energy Reorganization Act of 1974, as amended. Reports must be submitted in accordance with \u00a7 53.040.\n\n(b) [Reserved]\n\n(c) Records that are required by the regulations in this part, by license condition, or by technical specifications must be retained for the period specified by the appropriate regulation, license condition, or technical specification. If a retention period is not otherwise specified, these records must be retained until the Commission terminates the facility license or, in the case of an early site permit, until the permit expires.\n\n(d)(1) Records which must be retained under this part may be the original or a reproduced copy or a microform if the reproduced copy or microform is duly authenticated by authorized personnel and the microform is capable of producing a clear and legible copy after storage for the period specified by Commission regulations. The record may also be stored in electronic media with the capability of producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee must maintain adequate safeguards against tampering with, and loss of records.\n\n(2) If there is a conflict between the Commission's regulations in this part, license condition, or technical specification, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified in the regulations in this part for such records shall apply unless the Commission, under \u00a7 53.080 of this part, has granted a specific exemption from the record retention requirements in the regulations in this part.\n\n(e) Each licensee must notify the Commission as specified in \u00a7 53.040 of this part, of successfully completing power ascension testing or startup testing as applicable within 30 calendar days of completing the testing."], ["10:10:2.0.1.1.3.10.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1630 Immediate notification requirements for operating commercial nuclear plants.", "NRC", "", "", "", "(a)  General requirements. \n 1 \n   (1) Each holder of an OL under this part or a COL under this part after the Commission makes the finding under \u00a7 53.1452(g), must notify the NRC Headquarters Operations Center via the Emergency Notification System (ENS) of\u2014\n\n1  Other requirements for immediate notification of the NRC by licensed operating commercial nuclear plants are contained elsewhere in this chapter, in particular \u00a7\u00a7 20.1906, 20.2202, 72.216, 73.77, and 73.1200 of this chapter.\n\n(i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan; or\n\n(ii) Those non-emergency events specified in paragraph (b) of this section that occurred within 3 years of the date of discovery.\n\n(2) If the ENS is inoperative, the licensee must make the required notifications via commercial telephone service, other dedicated telephone system, or any other method which will ensure that a report is made as soon as practical to the NRC Headquarters Operations Center at the numbers specified in appendix A to part 73 of this chapter.\n\n(3) The licensee must notify the NRC immediately after notification of the appropriate State or local agencies and not later than 1 hour after the time the licensee declares one of the Emergency Classes.\n\n(4) The licensee must activate the data links with the NRC as specified in their emergency plans after declaring an Emergency Class for events of actual or potential substantial degradation of plant safety or security, probable risk to site personnel life, or site equipment damage caused by hostile action. The data links may also be activated by the licensee during emergency drills or exercises if the licensee's computer system has the capability to transmit the exercise data.\n\n(5) When making a report under paragraph (a)(1) of this section, the licensee must identify\u2014\n\n(i) The Emergency Class declared; or\n\n(ii) Paragraph (b)(1), \u201cOne-hour reports,\u201d paragraph (b)(2), \u201cFour-hour reports,\u201d or paragraph (b)(3), \u201cEight-hour reports,\u201d as the paragraph of this section requiring notification of the non-emergency event.\n\n(6) In lieu of submitting a report required under paragraph (b)(2) or (b)(3) of this section through the Emergency Notification System, the licensee may submit the report using other methods, provided the licensee submits the report to the NRC Headquarters Operations Center within the required timeframe and confirms receipt of the report by the NRC.\n\n(b)  Non-emergency events \u2014(1)  One-hour reports.  If not reported as a declaration of an Emergency Class under paragraph (a) of this section, the licensee must notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plant's Technical Specifications authorized under \u00a7 53.740(h) of this part.\n\n(2)  Four-hour reports.  If not reported under paragraphs (a) or (b)(1) of this section, the licensee must notify the NRC as soon as practical, and in all cases within 4 hours of the occurrence of any of the following:\n\n(i) The initiation of any commercial nuclear plant shutdown required by the plant's Technical Specifications.\n\n(ii) Any event or condition that results in actuation of the reactor protection system when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.\n\n(iii) [Reserved]\n\n(iv) Any event or condition that results in an unplanned movement of, change of state in, or chemical interaction involving a significant amount of radioactive material within the commercial nuclear plant.\n\n(v) [Reserved]\n\n(3)  Eight-hour reports.  If not reported under paragraphs (a), (b)(1), or (b)(2) of this section, the licensee must notify the NRC as soon as practical and in all cases within 8 hours of the occurrence of any of the following:\n\n(i) Any event or condition that results in\u2014\n\n(A) The condition of the commercial nuclear plant, including its principal safety barriers, being seriously degraded; or\n\n(B) The commercial nuclear plant being in a condition not analyzed under \u00a7 53.450 that significantly degrades plant safety.\n\n(ii)-(iv) [Reserved]\n\n(v) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability ( e.g.,  significant portion of control room indication, ENS, or offsite notification system).\n\n(c)  Follow-up notification:  With respect to the notifications made under paragraphs (a) and (b) of this section, in addition to making the required initial notification, each licensee, must during the course of the event\u2014\n\n(1) Immediately report:\n\n(i) Any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the Emergency Classes, if such a declaration has not been previously made; or\n\n(ii) Any change from one Emergency Class to another; or\n\n(iii) A termination of the Emergency Class.\n\n(2) Immediately report:\n\n(i) The results of ensuing evaluations or assessments of plant conditions,\n\n(ii) The effectiveness of response or protective measures taken, and\n\n(iii) Important information related to plant behavior that is not understood.\n\n(3) Maintain an open, continuous communication channel with the NRC Headquarters Operation Center upon request by the NRC."], ["10:10:2.0.1.1.3.10.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1640 Licensee event report system.", "NRC", "", "", "", "(a)  Reportable events.  (1) Each commercial nuclear plant licensee holding an OL under this part or a COL under this part after the Commission makes the finding under \u00a7 53.1452(g), must submit a Licensee Event Report (LER) for any event of the type described in this paragraph (a) within 60 days after discovery of the event. In the case of an invalid actuation reported under \u00a7 53.1640(a)(2), other than automatic reactor shutdown when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. Unless otherwise specified in this section, the licensee must report an event if it occurred within 3 years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event.\n\n(2) The licensee must report\u2014\n\n(i)(A) The completion of any commercial nuclear plant shutdown required by the plant's Technical Specifications.\n\n(B) Any operation or condition which was prohibited by the plant's Technical Specifications except when\u2014\n\n( 1 ) The Technical Specification is administrative in nature;\n\n( 2 ) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or\n\n( 3 ) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of the event.\n\n(C) Any deviation from the plant's Technical Specifications authorized under \u00a7 53.740(h).\n\n(ii) Any event or condition that resulted in\u2014\n\n(A) The condition of the commercial nuclear plant, including its principal safety barriers, being seriously degraded; or\n\n(B) The commercial nuclear plant being in a condition not analyzed under \u00a7 53.450 that significantly degrades plant safety.\n\n(iii) Any natural phenomena or other external condition that posed an actual threat to the safety of the commercial nuclear plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the commercial nuclear plant.\n\n(iv) Any event or condition that resulted in inadvertent operation of any structures, systems, and component classified as safety-related (SR) for an identified safety function under \u00a7 53.460 or the unplanned sole reliance on an SR system for those systems that are in constant operation, except when\u2014\n\n(A) The actuation resulted from and was part of a pre-planned sequence during testing; or\n\n(B) The actuation was invalid and\u2014\n\n( 1 ) Occurred while the system was properly removed from service; or\n\n( 2 ) Occurred after the safety function had been already completed.\n\n(v) Any event or condition that could have prevented the fulfillment of the safety functions identified under \u00a7 53.230.\n\n(vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (a)(2)(v) of this section if any other equipment was operable and available to perform the required safety function.\n\n(vii)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of any of the safety functions identified under \u00a7 53.230.\n\n(B) Events covered in paragraph (a)(2)(vii)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (a)(2)(vii)(A) of this section if the event results from\u2014\n\n( 1 ) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or\n\n( 2 ) Normal and expected wear or degradation.\n\n(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1-hour, resulted in airborne radionuclide concentrations in an unrestricted area that exceeds 20 times the applicable concentration limits specified in appendix B to 10 CFR part 20, table 2, column 1.\n\n(B) Any liquid effluent release that, when averaged over a time period of 1-hour, exceeds 20 times the applicable concentrations specified in appendix B to 10 CFR part 20, table 2, column 2, at the point of entry into the receiving waters ( i.e.,  unrestricted area) for all radionuclides except tritium and dissolved noble gases.\n\n(ix) Any event that posed an actual threat to the safety of the commercial nuclear plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the plant, including fires, toxic gas releases, or radioactive releases.\n\n(b)  Contents.  The LER must contain\u2014\n\n(1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence.\n\n(2)(i) A clear, specific narrative description of what occurred so that knowledgeable readers conversant with the design of commercial nuclear plants, but not familiar with the details of a particular plant, can understand the complete event.\n\n(ii) The narrative description must include the following specific information as appropriate for the particular event:\n\n(A) Plant operating conditions before the event.\n\n(B) Status of systems, structures, or components that were inoperable at the start of the event and that contributed to the event.\n\n(C) Dates and approximate time of the occurrences.\n\n(D) The cause of each component or system failure or personnel error, if known.\n\n(E) The failure mode, mechanism, and effect of each failed component, if known.\n\n(F) [Reserved]\n\n(G) For failures of components with multiple functions, include a list of systems or secondary functions that were also affected.\n\n(H) For failure that rendered a component or system classified as SR or non-safety-related but safety-significant inoperable, an estimate of the elapsed time from the discovery of the failure until the component or system was returned to service.\n\n(I) The method of discovery of each component or system failure or procedural error.\n\n(J) For each human performance related root cause, the licensee must discuss the cause(s) and circumstances.\n\n(K) Automatically and manually initiated safety system responses.\n\n(L) The manufacturer and model number (or other identification) of each component that failed during the event.\n\n(3) An assessment of the safety consequences and implications of the event. This assessment must include\u2014\n\n(i) The availability of systems or components that could have performed the same function as the components and systems that failed during the event, and\n\n(ii) For events that occurred when the reactor was shut down, the availability of systems or components that are needed to shut down the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident.\n\n(4) A description of any corrective actions planned as a result of the event, including those to reduce the probability of similar events occurring in the future.\n\n(5) Reference to any previous similar events at the same plant that are known to the licensee.\n\n(6) The name and contact information of a person within the licensee's organization who is knowledgeable about the event and can provide additional information concerning the event and the plant's characteristics.\n\n(c)  Supplemental information.  The Commission may require the licensee to submit specific additional information beyond that required by paragraph (b) of this section if the Commission finds that supplemental material is necessary for complete understanding of an unusually complex or significant event. These requests for supplemental information will be made in writing and the licensee must submit, as specified in \u00a7 53.040, the requested information as a supplement to the initial LER.\n\n(d)  Submission of reports.  Licensee Event Reports must be prepared on NRC Form 366 and submitted to the NRC, as specified in \u00a7 53.040.\n\n(e)  Report legibility.  The reports and copies that licensees are required to submit to the Commission under the provisions of this section must be of sufficient quality to permit legible reproduction and micrographic processing."], ["10:10:2.0.1.1.3.10.62.6", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1645 Reports of radiation exposure to members of the public.", "NRC", "", "", "", "(a) Each holder of an OL, and each holder of a COL after the Commission has made the finding under \u00a7 53.1452(g), must submit radiological reports as required by 10 CFR part 20, as well as an Annual Radioactive Effluent Release Report and an Annual Radiological Environmental Operating Report. The Annual Radioactive Effluent Release Report must specify the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents and an estimate of the dose received by the maximally exposed member of the public in an unrestricted area from effluents and direct radiation from contained sources during the previous calendar year. The Annual Radiological Environmental Operating Report must provide data on measurable levels of radiation and radioactive materials in the environment, must include an evaluation of the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure, and must include the results of environmental monitoring during the previous calendar year. These reports must also include any other information as may be required by the Commission to estimate maximum potential annual radiation doses to the public. The reports must be submitted as specified in \u00a7 53.040 by May 15 of each successive year. If the total effective dose equivalent to members of the public in unrestricted areas during the reporting period is greater than the design objectives established under \u00a7 53.425, the report must specify the causes for exceeding the design objective and describe any corrective actions. On the basis of these reports and any additional information the Commission may obtain from the licensee or others, the Commission may require the licensee to take action as the Commission deems appropriate.\n\n(b) If during any calendar quarter the radiation exposure to a member of the public in the unrestricted areas, calculated on the same basis as the respective design objective exposure, exceeds one-half of the annual design objective exposure, the licensee must submit a report as specified in \u00a7 53.040. The report shall specify the causes for exceeding one-half the annual design objective exposure in a quarter and describe corrective actions that the licensee will take to maintain radiation exposure to levels within the design objectives for the remainder of the year. The report shall be submitted within 30 days from the end of the quarter when one-half of the annual design objective exposure was exceeded."], ["10:10:2.0.1.1.3.10.62.7", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1650 Facility information and verification.", "NRC", "", "", "", "(a) In response to a written request by the Commission, each applicant for a CP or license and each recipient of a CP or a license must submit facility information, as described in \u00a7 75.10 of this chapter, on International Atomic Energy Agency (IAEA) Design Information Questionnaire forms and site information on DOC/NRC Form AP-A and associated forms;\n\n(b) As required by the Additional Protocol, must submit location information described in \u00a7 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms; and\n\n(c) Must permit verification thereof by the IAEA and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in part 75 of this chapter."], ["10:10:2.0.1.1.3.10.62.8", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1660 Financial requirements.", "NRC", "", "", "", "Sections 53.1670 through 53.1700 set out the requirements and procedures related to financial qualifications and related reporting requirements."], ["10:10:2.0.1.1.3.10.62.9", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "J", "Subpart J\u2014Reporting and Other Administrative Requirements", "", "\u00a7 53.1670 Financial qualifications.", "NRC", "", "", "", "Except for an electric utility applicant for a license to operate a commercial nuclear plant, an applicant for a CP, OL, or COL under this part must appear to be financially qualified for the activities for which the permit or license is sought."], ["10:10:2.0.1.1.3.12.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "M", "Subpart M\u2014Enforcement", "", "\u00a7 53.9000 Violations.", "NRC", "", "", "", "(a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of\u2014\n\n(1) The Atomic Energy Act of 1954, as amended (the Act);\n\n(2) Title II of the Energy Reorganization Act of 1974, as amended; or\n\n(3) A regulation or order issued under those Acts.\n\n(b) The Commission may obtain a court order for the payment of a civil penalty imposed under Section 234 of the Act:\n\n(1) For violations of\u2014\n\n(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Act;\n\n(ii) Section 206 of the Energy Reorganization Act of 1974, as amended;\n\n(iii) Any rule, regulation, or order issued under the sections specified in paragraph (b)(1)(i) of this section;\n\n(iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.\n\n(2) For any violation for which a license may be revoked under section 186 of the Act."], ["10:10:2.0.1.1.3.12.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "M", "Subpart M\u2014Enforcement", "", "\u00a7 53.9010 Criminal penalties.", "NRC", "", "", "", "(a) Section 223 of the Act provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 53 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.\n\n(b) The regulations in 10 CFR part 53 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: \u00a7\u00a7 53.000, 53.015, 53.020, 53.040, 53.080, 53.090, 53.100, 53.110, 53.120, 53.600, 53.725, 53.726, 53.735, 53.760, 53.775, 53.790, 53.795, 53.820, 53.910, 53.1000, 53.1050, 53.1100, 53.1103, 53.1106, 53.1109, 53.1112, 53.1115, 53.1118, 53.1120, 53.1121, 53.1124, 53.1140, 53.1144, 53.1146, 53.1149, 53.1155, 53.1158, 53.1164, 53.1170, 53.1173, 53.1176, 53.1179, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210, 53.1212, 53.1215, 53.1218, 53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1242, 53.1245, 53.1248, 53.1251, 53.1254, 52.1257, 52.1260, 53.1263, 53.1270, 53.1276, 53.1279, 53.1282, 53.1285, 53.1286, 53.1287, 53.1288, 53.1291, 53.1293, 53.1295, 53.1300, 53.1306, 53.1309, 53.1312, 53.1315, 53.1318, 53.1324, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1375, 53.1381, 53.1384, 53.1387, 53.1390, 53.1396, 53.1401, 53.1405, 53.1410, 53.1416, 53.1419, 53.1422, 53.1425, 53.1431, 53.1437, 53.1440, 53.1443, 53.1452, 53.1455, 53.1456, 53.1458, 53.1461, 53.1470, 53.1500, 53.1510, 53.1515, 53.1520, 53.1525, 53.1530, 53.1535, 53.1540, 53.1560, 53.1585, 53.1590, 53.1595, 53.1600, 53.1660, 53.1670, 53.1700, 53.1710, 53.1730, 53.9000, 53.9010."], ["10:10:2.0.1.1.3.2.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.210 Safety criteria for design-basis accidents.", "NRC", "", "", "", "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 \u00a7 53.240 demonstrate the following:\n\n(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\n\n(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.\n 1\n\n1  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."], ["10:10:2.0.1.1.3.2.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.220 Safety criteria for licensing-basis events other than design-basis accidents.", "NRC", "", "", "", "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 \u00a7 53.240 demonstrate the following:\n\n(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 \u00a7\u00a7 53.240 and 53.450(e), and provide measures for defense in depth in accordance with \u00a7 53.250; and\n\n(b) The analysis of risks to public health and safety resulting from LBEs other than DBAs under \u00a7 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."], ["10:10:2.0.1.1.3.2.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.230 Safety functions.", "NRC", "", "", "", "(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.\n\n(b) Additional safety functions needed to support the retention of radioactive materials during LBEs\u2014such as controlling reactivity, heat generation, heat removal, and chemical interactions\u2014must be identified for each commercial nuclear plant.\n\n(c) The primary and additional safety functions are required to satisfy the safety criteria defined in \u00a7\u00a7 53.210 and 53.220 and must be fulfilled by the design features, human actions, and programmatic controls specified throughout this part."], ["10:10:2.0.1.1.3.2.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.240 Licensing-basis events.", "NRC", "", "", "", "(a) Licensing-basis events must be identified for each commercial nuclear plant and analyzed under \u00a7 53.450 to demonstrate that the safety requirements in this subpart have been satisfied.\n\n(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.\n\n(c) The analysis of LBEs must\u2014\n\n(1) Include analysis of one or more DBAs under \u00a7 53.450(f);\n\n(2) Confirm the adequacy of design features and programmatic controls needed to satisfy the safety criteria defined in \u00a7\u00a7 53.210 and 53.220, and\n\n(3) Establish related functional requirements for plant SSCs, personnel, and programs."], ["10:10:2.0.1.1.3.2.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.250 Defense in depth.", "NRC", "", "", "", "(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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.2.62.6", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.260 Normal operations.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.2.62.7", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "B", "Subpart B\u2014Technology-Inclusive Safety Requirements", "", "\u00a7 53.270 Protection of plant workers.", "NRC", "", "", "", "Holders of licenses to operate commercial nuclear plants under this part must control occupational doses to meet the requirements in 10 CFR part 20."], ["10:10:2.0.1.1.3.3.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.400 Design features for licensing-basis events.", "NRC", "", "", "", "(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 \u00a7\u00a7 53.210 and 53.220.\n\n(b) Design features must ensure that the safety functions identified in \u00a7 53.230 are fulfilled during licensing-basis events (LBEs)."], ["10:10:2.0.1.1.3.3.62.10", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.470 [Reserved]", "NRC", "", "", "", ""], ["10:10:2.0.1.1.3.3.62.11", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.480 Earthquake engineering.", "NRC", "", "", "", "(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 \u00a7 53.230.\n\n(b)  Definitions.  As used in this section\u2014\n\nDesign-Basis Ground Motions (DBGMs)  are the vibratory ground motions for which certain SSCs must be designed to remain functional.\n\nOperating 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 \u00a7 53.720.\n\nResponse 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.\n\nSurface deformation  is the distortion of geologic strata on or near the ground surface that occurs because of tectonic forces that result from earthquakes.\n\n(c)  Design considerations \u2014(1)  Design-Basis Ground Motions.  (i) The DBGMs must be derived from the Site Ground Motion Response Spectra developed in accordance with \u00a7 53.510(c), by taking into consideration the functional design criteria of SSCs in accordance with \u00a7\u00a7 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.\n\n(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:\n\n(A) Structures, systems, and components for which functional design criteria are established in accordance with \u00a7 53.410 or \u00a7 53.420; and\n\n(B) Structures, systems, and components classified as SR or NSRSS commensurate with safety significance in accordance with \u00a7 53.460.\n\n(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.\n\n(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.\n\n(v) The SSCs fulfilling the safety functions required by \u00a7 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.\n\n(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.\n\n(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.\n\n(3) [Reserved]\n\n(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.\n\n(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 \u00a7 53.460 will remain functional.\n\n(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.\n\n(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.\n\n(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.\n\n(f)  Analysis.  The analyses required by \u00a7 53.450 must address seismic hazards and related SSC responses in determining that the safety criteria defined in \u00a7 53.220 will be met.\n\n(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 \u00a7 53.220 and to maintain consistency with analyses required by \u00a7 53.450 when accounting for the site-specific frequencies and magnitudes of earthquakes for a commercial nuclear plant."], ["10:10:2.0.1.1.3.3.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.410 Functional design criteria for design-basis accidents.", "NRC", "", "", "", "(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 \u00a7 53.210.\n\n(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 \u00a7 53.210, and to maintain consistency with analyses required by \u00a7 53.450(f)."], ["10:10:2.0.1.1.3.3.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.415 Protection against external hazards.", "NRC", "", "", "", "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 \u00a7 53.510 without losing the capability to perform the safety functions identified under \u00a7 53.230. Specific requirements for earthquake engineering are included in \u00a7 53.480."], ["10:10:2.0.1.1.3.3.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.420 Functional design criteria for licensing-basis events other than design-basis accidents.", "NRC", "", "", "", "(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\u2014\n\n(1) The safety criteria in \u00a7 53.220; and\n\n(2) The evaluation criteria in \u00a7 53.450(e).\n\n(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\u2014\n\n(1) The safety criteria in \u00a7 53.220; and\n\n(2) The evaluation criteria in \u00a7 53.450(e)."], ["10:10:2.0.1.1.3.3.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.425 Design features and functional design criteria for normal operations.", "NRC", "", "", "", "(a) Design features must be provided for each commercial nuclear plant to support the Radiation Protection Program required in \u00a7 53.850.\n\n(b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with \u00a7 53.850.\n\n(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."], ["10:10:2.0.1.1.3.3.62.6", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.430 Design features and functional design criteria for protection of plant workers.", "NRC", "", "", "", "(a) Design features must be provided for each commercial nuclear plant such that, when combined with corresponding programmatic controls, the requirements in \u00a7 53.270 can be met.\n\n(b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with \u00a7 53.270."], ["10:10:2.0.1.1.3.3.62.7", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.440 Design requirements.", "NRC", "", "", "", "(a)(1) Analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof must demonstrate that each design feature required by \u00a7 53.400 meets the defined functional design criteria required by \u00a7\u00a7 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 \u00a7 53.400 must function throughout the plant's lifetime.\n\n(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.\n\n(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).\n\n(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 \u00a7 53.460.\n\n(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 \u00a7 53.870.\n\n(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.\n\n(2) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs.\n\n(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.\n\n(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 \u00a7 53.230.\n\n(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.\n\n(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 \u00a7 53.240.\n\n(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 \u00a7 53.240.\n\n(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.\n\n(j) [Reserved]\n\n(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.\n\n(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 \u00a7 20.1406 of this chapter.\n\n(m)(1) Each commercial nuclear plant must include criticality monitoring capabilities meeting the requirements of either \u00a7 70.24 of this chapter or paragraph (m)(2) of this section.\n\n(2) In lieu of maintaining a monitoring system capable of detecting criticality as described in \u00a7 70.24 of this chapter, criticality accident requirements may be satisfied by\u2014\n\n(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\n\n(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.\n\n(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\n(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.\n\n(2) The design must provide for the capabilities described in \u00a7 53.730(b) to ensure the plant staff are able to monitor plant conditions and respond to events.\n\n(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 \u00a7 53.730(c).\n\n(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."], ["10:10:2.0.1.1.3.3.62.8", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.450 Analysis requirements.", "NRC", "", "", "", "(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 \u00a7 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of \u00a7 53.220.\n\n(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\u2014\n\n(1) Inform the selection of the LBEs, as described in \u00a7 53.240, which must be considered in the design to determine compliance with the safety criteria in subpart B of this part.\n\n(2) Inform the classification of SSCs according to their safety significance in accordance with \u00a7 53.460 and to identify the environmental conditions under which the SSCs and operating staff must perform their safety functions.\n\n(3) Evaluate the adequacy of defense-in-depth measures required in accordance with \u00a7 53.250.\n\n(4) Identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment.\n\n(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.\n\n(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 \u00a7\u00a7 53.220 and the analyses of licensing-basis events other than design-basis accidents (DBAs) under \u00a7 53.450(e).\n\n(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 \u00a7 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.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 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 \u00a7 53.220 and provide defense in depth as required by \u00a7 53.250.\n\n(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.\n\n(f)  Analysis of design-basis accidents.  (1) The analysis of LBEs required by \u00a7 53.240 must include analysis of DBAs that address possible challenges to the safety functions identified under \u00a7 53.230. The events selected as DBAs must be those that, if not terminated, have the potential for exceeding the safety criteria in \u00a7 53.210.\n\n(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 \u00a7 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 \u00a7 53.230.\n\n(3) The analysis must conservatively demonstrate compliance with the safety criteria in \u00a7 53.210.\n\n(g)  Other required analyses.  Analyses must be performed to assess\u2014\n\n(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\u2014\n\n(i) Prevent equipment from fulfilling the safety functions identified in accordance with \u00a7 53.230; or\n\n(ii) Challenge the safety criteria in \u00a7\u00a7 53.210 and 53.220.\n\n(2) [Reserved]\n\n(3)  Dose to members of the public.  Measures taken under \u00a7 53.425, including estimating\u2014\n\n(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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.3.62.9", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "C", "Subpart C\u2014Design and Analysis Requirements", "", "\u00a7 53.460 Safety categorization and special treatments.", "NRC", "", "", "", "(a) Structures, systems, and components must be classified according to their safety significance. The SSC categories must include \u201cSafety-Related,\u201d \u201cNon-Safety-Related but Safety-Significant,\u201d and \u201cNon-Safety-Significant,\u201d as defined in subpart A of this part.\n\n(b) For SR and NSRSS SSCs, the conditions under which they must perform their safety function in \u00a7 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 \u00a7 53.450 to demonstrate meeting the safety criteria in \u00a7\u00a7 53.210 and 53.220.\n\n(1) The special treatments for SR SSCs must include meeting the applicable quality assurance requirements from appendix B of part 50 of this chapter.\n\n(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.\n\n(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 \u00a7 53.450 to demonstrate meeting the applicable criteria in \u00a7\u00a7 53.210, 53.220, and 53.450(e)."], ["10:10:2.0.1.1.3.4.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "D", "Subpart D\u2014Siting Requirements", "", "\u00a7 53.500 General siting and siting assessment.", "NRC", "", "", "", "The purpose of this subpart and the specific requirements therein is to ensure that:\n\n(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 \u00a7\u00a7 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 \u00a7\u00a7 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.\n\n(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 \u00a7 53.460, including, where applicable, the quality assurance requirements from appendix B of part 50 of this chapter."], ["10:10:2.0.1.1.3.4.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "D", "Subpart D\u2014Siting Requirements", "", "\u00a7 53.510 External hazards.", "NRC", "", "", "", "(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.\n\n(b)  Definitions.  As used in this section, the following terms mean:\n\nGeological Siting Factors  are geological and seismic factors that may affect the design and operation of the proposed commercial nuclear plant.\n\nGround 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 \u00a7 53.480.\n\nProbabilistic 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.\n\n(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.\n\n(d)  Geologic and seismic siting factors.  The geologic and seismic siting factors considered for design under \u00a7\u00a7 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."], ["10:10:2.0.1.1.3.4.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "D", "Subpart D\u2014Siting Requirements", "", "\u00a7 53.520 Site characteristics.", "NRC", "", "", "", "Site characteristics that might contribute to the initiation, progression, or consequences of LBEs analyzed under \u00a7 53.450 must be identified, assessed, and considered in the design and analyses required by subpart C of this part."], ["10:10:2.0.1.1.3.4.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "D", "Subpart D\u2014Siting Requirements", "", "\u00a7 53.530 Population-related considerations.", "NRC", "", "", "", "Every site must have an exclusion area, a low-population zone, and a population center distance as defined in \u00a7 53.020.\n\n(a) The offsite radiological consequences estimated by the analyses required by \u00a7 53.450(f) must be used to confirm that\u2014\n\n(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.\n\n(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.\n\n(b) The reactor site must either:\n\n(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\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.4.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "D", "Subpart D\u2014Siting Requirements", "", "\u00a7 53.540 Siting interfaces.", "NRC", "", "", "", "Site characteristics must be addressed by the design features, programmatic controls, and supporting analyses used to demonstrate that the safety criteria in \u00a7\u00a7 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."], ["10:10:2.0.1.1.3.5.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "E", "Subpart E\u2014Construction and Manufacturing Requirements", "", "\u00a7 53.600 Construction and manufacturing\u2014scope and purpose.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.5.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "E", "Subpart E\u2014Construction and Manufacturing Requirements", "", "\u00a7 53.605 Reporting of defects and noncompliance.", "NRC", "", "", "", "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 \u00a7 53.1452(g).\n\n(a)  Definitions.  The definitions in \u00a7 21.3 of this chapter apply to this section.\n\n(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.\n\n(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.\n\n(c)  Procedures.  The holder of a CP, COL, or ML subject to this section must adopt appropriate procedures to\u2014\n\n(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.\n\n(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.\n\n(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\u2014\n\n(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;\n\n(ii) Contains a defect; or\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(e)  Notification\u2014timing and where sent.  The notification required by paragraph (d) of this section must consist of\u2014\n\n(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.\n\n(2) Written notification submitted to the NRC Document Control Desk by an appropriate method listed in \u00a7 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.\n\n(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.\n\n(1) Name and address of the individual or individuals informing the Commission.\n\n(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.\n\n(3) Identification of the firm constructing or manufacturing the facility or supplying the basic component which fails to comply or contains a defect.\n\n(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.\n\n(5) The date on which the information of a defect or failure to comply was obtained.\n\n(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.\n\n(7) In the case of a completed reactor manufactured under this part, the entities to which the reactor was supplied.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 73.1205 of this chapter, under this section, or under \u00a7 53.1640.\n\n(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\u2014\n\n(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.\n\n(2) Retain records of evaluations of all deviations and failures to comply under paragraph (c)(1) of this section for the longest of\u2014\n\n(i) Ten years from the date of the evaluation;\n\n(ii) Five years from the date that an early site permit is referenced in an application for a COL; or\n\n(iii) Five years from the date of delivery of a manufactured reactor.\n\n(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;\n\n(4) Suppliers of basic components must retain records of\u2014\n\n(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;\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.5.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "E", "Subpart E\u2014Construction and Manufacturing Requirements", "", "\u00a7 53.610 Construction.", "NRC", "", "", "", "(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:\n\n(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.\n\n(2) An organization, headed by qualified personnel, responsible for managing, controlling, and evaluating the adequacy of the construction activities.\n\n(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.\n\n(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.\n\n(5) A fitness-for-duty program, under 10 CFR part 26.\n\n(6)(i) A QAP meeting the requirements of appendix B of part 50 of this chapter as required by \u00a7 53.460(b).\n\n(ii) Appropriate programmatic controls to provide special treatment for non-safety-related but safety-significant structures, systems, and components (SSCs).\n\n(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.\n\n(8) An information security program in accordance with \u00a7\u00a7 73.21, 73.22, and 73.23 of this chapter, as applicable.\n\n(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 \u00a7 53.1130, or an LWA under this part.\n\n(1) Licensees must satisfy the following requirements:\n\n(i) As appropriate, considering the types and quantities of radioactive materials being brought onto the site\u2014\n\n(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 \u00a7 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.\n\n(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.\n\n(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 \u00a7 30.18 of this chapter).\n\n(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.\n\n(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.\n\n(iv) For LWA holders, a plan must be developed for redress of activities performed under the LWA should one of the following situations arise:\n\n(A) LWA work activities are terminated by the holder of the LWA;\n\n(B) The LWA is revoked by the NRC; or\n\n(C) The Commission denies the associated CP or COL application.\n\n(2)(i) Onsite fresh fuel must be protected and stored in compliance with \u00a7 73.67 of this chapter.\n\n(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 \u00a7 53.620(d)(1)), a cybersecurity program that meets the requirements of \u00a7 73.54 or \u00a7 73.110 of this chapter, a physical security program that meets the requirements of \u00a7 73.55 or \u00a7 73.100 of this chapter, and an access authorization program that meets the requirements of \u00a7 73.56 or \u00a7 73.120 of this chapter must be established, as applicable.\n\n(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 \u00a7 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.\n\n(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.\n\n(2) The post-construction acceptance process must address the inspections, tests, analyses, and acceptance criteria specified in the COL under \u00a7 53.1440 or the equivalent verifications needed to support the issuance of an operating license under \u00a7 53.1387."], ["10:10:2.0.1.1.3.5.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "E", "Subpart E\u2014Construction and Manufacturing Requirements", "", "\u00a7 53.620 Manufacturing.", "NRC", "", "", "", "(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:\n\n(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.\n\n(2) An organizational and management structure responsible for managing, controlling, and evaluating the adequacy of the reactor design and manufacturing activities.\n\n(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.\n\n(4) A program to evaluate the applicability of other national and international design and manufacturing experience to the planned and ongoing manufacturing activities.\n\n(5) A fitness-for-duty program, in accordance with 10 CFR part 26.\n\n(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.\n\n(ii) Appropriate programmatic controls to provide special treatment measures for non-safety-related but safety-significant SSCs.\n\n(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.\n\n(8) An information security program in accordance with \u00a7\u00a7 73.21, 73.22 and 73.23 of this chapter, as applicable.\n\n(b)  Manufacturing activities.  Holders of MLs must satisfy the following requirements:\n\n(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.\n\n(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.\n\n(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.\n\n(c)  Control of radioactive materials.  As appropriate considering the types and quantities of radioactive materials being brought into the manufacturing facility\u2014\n\n(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.\n\n(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 \u00a7 30.18 of this chapter).\n\n(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 \u00a7 30.18 of this chapter).\n\n(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 \u00a7 30.18 of this chapter).\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 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.\n\n(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.\n\n(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:\n\n(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.\n\n(B) The physical security program must be designed to prevent unintended and uncontrolled criticality events.\n\n(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.\n\n(D) All holders of a part 70 license that authorizes loading of fresh fuel into a manufactured reactor must perform the screening required in \u00a7 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.\n\n(ii) [Reserved]\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 53.610.\n\n(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."], ["10:10:2.0.1.1.3.6.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.700 Operational objectives.", "NRC", "", "", "", "The purpose of this subpart and the specific requirements herein is to ensure that:\n\n(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:\n\n(1) Under \u00a7 53.710 through \u00a7 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 \u00a7 53.230 will be performed if called upon during licensing-basis events (LBEs).\n\n(2) Under \u00a7 53.725 through \u00a7 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 \u00a7 53.230.\n\n(3) Under \u00a7 53.845 through \u00a7 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 \u00a7 53.230 will be performed if called upon during normal operations and LBEs.\n\n(b) [Reserved]"], ["10:10:2.0.1.1.3.6.62.10", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.740 Facility licensee requirements\u2014general.", "NRC", "", "", "", "(a) Facility licensees must demonstrate compliance with the requirements of either \u00a7\u00a7 53.760 through 53.795 for interaction-dependent-mitigation facilities or \u00a7\u00a7 53.800 through 53.820 for self-reliant-mitigation facilities.\n\n(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 \u00a7 53.1070. The approved staffing plan is subject to the requirements of \u00a7 53.1565.\n\n(c) Except as provided under \u00a7 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.\n\n(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 \u00a7 53.1070 must designate senior operators to be responsible for supervising the licensed activities of operators.\n\n(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.\n\n(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:\n\n(i) The actuation of an automatic protection system that utilizes setpoints more conservative than those otherwise credited for the purposes of reactor protection; or\n\n(ii) An automated control system; or\n\n(iii) An operator or senior operator or a generally licensed reactor operator, as appropriate.\n\n(2) The provisions of paragraph (e) of this section do not apply during load following operations.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 53.1070 by a certified fuel handler, senior operator, or generally licensed reactor operator, as applicable, prior to taking the action."], ["10:10:2.0.1.1.3.6.62.11", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.745 Operator license requirements.", "NRC", "", "", "", "A person must be authorized by 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."], ["10:10:2.0.1.1.3.6.62.12", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.760 Operator licensing.", "NRC", "", "", "", "(a)  Applicability.  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 \u00a7 53.1070.\n\n(b) [Reserved]"], ["10:10:2.0.1.1.3.6.62.13", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.765 Medical requirements.", "NRC", "", "", "", "(a) An applicant for an operator or senior operator license must have a medical examination by a physician. An operator or senior operator must have a medical examination by a physician every 2 years.\n\n(b) To certify the medical fitness of an applicant for an operator or senior operator license, an authorized representative of the facility licensee must complete and sign NRC Form 396, \u201cCertification of Medical Examination by Facility Licensee,\u201d which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-7232, or by visiting the NRC's website at  https://www.nrc.gov  and selecting forms from the index found on the home page, or by other means provided by the NRC.\n\n(1) NRC Form 396 must certify that a physician has conducted the medical examination of the applicant as required in paragraph (a) of this section.\n\n(2) When the medical certification requests a conditional license based on medical evidence, the medical evidence must be submitted on NRC Form 396 to the Commission to enable the Commission to make a determination in accordance with \u00a7 53.775(b).\n\n(c) The facility licensee must document and maintain the results of medical qualifications data, test results, and each operator's or senior operator's medical history for the current license period and provide the documentation to the Commission upon request. The facility licensee must retain this documentation while an individual performs the functions of an operator or senior operator."], ["10:10:2.0.1.1.3.6.62.14", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.770 Incapacitation because of disability or illness.", "NRC", "", "", "", "If, during the term of the operator or senior operator license, the licensee develops a permanent physical or mental condition that causes the licensee to fail to demonstrate compliance with the requirements of \u00a7 53.775(b)(1)(i), the facility licensee must notify the Commission within 30 days of learning of the diagnosis. For conditions for which a conditional license (as described in \u00a7 53.775(b)) is requested, the facility licensee must provide medical certification on NRC Form 396 to the Commission (as described in \u00a7 53.765(b))."], ["10:10:2.0.1.1.3.6.62.15", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.775 Applications for operators and senior operators.", "NRC", "", "", "", "(a)  How to apply.  (1) The applicant for an operator or senior operator license must\u2014\n\n(i) Complete NRC Form 398, \u201cPersonal Qualification Statement\u2014Licensee,\u201d which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-5877, or by visiting the NRC's website at  https://www.nrc.gov  and selecting forms from the index found on the home page, or by other means provided by the NRC;\n\n(ii) File an original of NRC Form 398, or an equivalent electronic submittal, together with the information required in paragraphs (a)(1)(iii) and (a)(1)(iv) of this section, with the appropriate Regional Administrator.\n\n(iii) Provide evidence that the applicant, as a trainee, has successfully demonstrated competence in manipulating the controls of either the facility for which a license is sought or a simulation facility that demonstrates compliance with the requirements of \u00a7 53.780(e). For operators applying for a senior operator license, certification that the operator has successfully operated the controls of the facility as an operator will be accepted; and\n\n(iv) Provide certification by the facility licensee of medical condition and general health on NRC Form 396, to comply with \u00a7 53.765.\n\n(2) The Commission may at any time after the application has been filed, and before the license has expired, require further information under oath or affirmation to enable it to determine whether to grant or deny the application or whether to revoke, modify, or suspend the license.\n\n(3) An applicant whose application has been denied because of a medical condition or their general health may submit a further medical report at any time as a supplement to the application.\n\n(4) Each application and statement must contain complete and accurate disclosure as to all matters required to be disclosed. The applicant must sign statements required by paragraphs (a)(1)(i) and (a)(1)(ii) of this section.\n\n(b)  Disposition of an initial application \u2014(1)  License approval.  The Commission will approve an initial application if it finds that the following criteria are met:\n\n(i)  Health.  The applicant's medical condition and general health will not adversely affect the performance of assigned operator or senior operator job duties or cause operational errors endangering public health and safety. The Commission will base its finding upon the certification by the facility licensee as detailed in \u00a7 53.765(b).\n\n(ii)  Examination.  The applicant has passed the requisite examination in accordance with \u00a7 53.780(b). The examination determines whether the applicant for an operator's or senior operator's license has learned to operate a facility competently and safely, and additionally, in the case of a senior operator, whether the applicant has learned to supervise the licensed activities of operators competently and safely.\n\n(2)  Conditional license.  If an applicant's general medical condition does not demonstrate compliance with the minimum standards under \u00a7 53.775(b)(1)(i), the Commission may approve the application and include conditions in the license to accommodate the medical condition. The Commission will consider the recommendations and supporting evidence of the facility licensee and of the examining physician (provided on NRC Form 396) in arriving at its decision.\n\n(c)  Re-applications.  (1) An applicant whose application for a license has been denied because of failure to pass the examination may file a new application. The application must be submitted on NRC Form 398 and include a statement signed by an authorized representative of the facility licensee by whom the applicant will be employed that states in detail the extent of the applicant's additional training and remediation since the denial and certifies that the applicant is ready for re-examination.\n\n(2) An applicant who has passed a portion of the examination and failed another may request in a new application on NRC Form 398 to be excused from re-examination on the portions of the examination that the applicant has passed. The Commission may in its discretion grant the request if it determines that sufficient justification is presented."], ["10:10:2.0.1.1.3.6.62.16", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.780 Training, examination, and proficiency program.", "NRC", "", "", "", "(a)  Operator licensing initial training program.  (1) A program that is based upon a systems approach to training, as defined by \u00a7 53.725(b), must be utilized for the training of applicants for operator and senior operator licenses. The program must ensure that applicants at the facility will possess the knowledge, skills, and abilities necessary to protect the public health and maintain those plant safety functions specific to the facility design. The program must be approved by the Commission prior to its use for training applicants, as described under \u00a7 53.730(g). The approved operator licensing initial training program is subject to the requirements of \u00a7 53.1565.\n\n(2) The facility licensee must maintain operator licensing initial training program records documenting the initial operator licensing training administered and completed by each applicant. The facility licensee must retain these records during the period in which any trainees subsequently remain licensed as operators or senior operators at the facility.\n\n(b)  Operator licensing initial examination program.  (1) The facility licensee must establish and implement an examination program for testing a representative sample of the knowledge, skills, and abilities needed to safely perform operator and senior operator duties, to include both the examination methods and criteria to be used to assess passing performance. The program must provide for valid and reliable examinations and be approved by the Commission prior to its use for examining applicants, as described under \u00a7 53.730(g). The approved operator licensing initial examination program is subject to the requirements of \u00a7 53.1565.\n\n(2) The facility licensee must submit prepared examinations to the Commission for review and approval in advance of their administration.\n\n(3) The Commission will either administer an approved examination or allow the facility licensee to administer the examination. The facility licensee must ensure that sufficient advance notification is provided to the Commission to either administer the examination or allow for a representative of the Commission to be afforded the opportunity to be present when the facility licensee administers the examination.\n\n(4) Graded examination documentation for each applicant must be provided to the Commission for review in making operator licensing decisions.\n\n(5) The facility licensee must maintain operator licensing initial examination program records documenting the participation of each operator and senior operator applicant in the initial examination. The records must contain copies of examinations administered, the answers given by the applicant, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an applicant exhibited deficiencies. The facility licensee must retain these records during the period in which the associated operators or senior operators remain licensed at the facility.\n\n(c)  Operator licensing requalification program.  (1) A program based upon a systems approach to training, as defined by \u00a7 53.725(b), must be utilized for the continuing training of operators and senior operators.\n\n(i) The program must ensure that operators and senior operators at the facility maintain the knowledge, skills, and abilities necessary to protect the public health and maintain those plant safety functions specific to the facility design. The program must be conducted for a continuous period not to exceed 24 months in duration.\n\n(ii) The program must be approved by the Commission prior to its use for continuing training, as described under \u00a7 53.730(g). The approved operator licensing requalification program is subject to the requirements of \u00a7 53.1565.\n\n(2) The following requirements apply to operator licensing requalification programs:\n\n(i) The facility licensee must propose a requalification examination program for testing, for each requalification period, a sample of the topics included under the systems approach to training, to include both the examination methods and criteria to be used to assess passing performance. The program must provide for valid and reliable examinations and be approved by the Commission prior to its use for examining operators and senior operators, as described under \u00a7 53.730(g). The approved requalification examination program is subject to the requirements of \u00a7 53.1565.\n\n(ii) The following requirements apply to the requalification examination program:\n\n(A) The facility licensee must make prepared requalification examinations available to the Commission for review.\n\n(B) The facility licensee must ensure that a representative of the Commission is afforded the opportunity to be present during requalification examination administration.\n\n(C) The facility licensee must ensure that each operator and senior operator is administered a complete requalification examination on a periodicity not to exceed 24 months. Additionally, the facility licensee must ensure that any licensed operator or senior licensed operator who either demonstrates unsatisfactory performance on, or fails to complete, this biennial requalification examination is removed from the performance of licensed operator and senior licensed operator duties until any necessary remedial training has been completed and a retake examination has been passed.\n\n(D) The facility licensee must promptly provide a summary of examination results to the NRC for each operator and senior operator following the completion of the requalification examination.\n\n(3) The facility licensee must maintain operator licensing requalification program records documenting the participation of each operator and senior operator in the requalification program. The records must contain copies of examinations administered, the answers given by the operator or senior operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an operator or senior operator exhibited deficiencies. The facility licensee must retain these records until the operator's or senior operator's license is renewed.\n\n(d)  Examination integrity.  Applicants, operators and senior operators, and facility licensees must not engage in any activity that compromises the integrity of any application or examination required by \u00a7\u00a7 53.760 through 53.795. The integrity of an examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, could have affected the consistent administration of the examination. This includes activities related to the preparation and certification of applications and all activities related to the preparation, administration, and grading of examinations required by \u00a7\u00a7 53.760 through 53.795.\n\n(e)  Simulation facilities.  (1) This section addresses the use of a simulation facility for the administration of examinations, for training, or to demonstrate compliance with experience requirements for applicants for operator and senior operator licenses.\n\n(2) Simulation facilities used for training purposes, for demonstrating compliance with experience requirements, or for the conduct of examinations under \u00a7 53.780(b) and (c) must demonstrate compliance with the following criteria as they relate to the facility licensee's reference plant:\n\n(i) The simulation facility must be of sufficient scope and fidelity for individuals to acquire and demonstrate the necessary knowledge, skills, and abilities to safely perform operator and senior operator duties.\n\n(ii) The simulation facility must utilize models relating to nuclear, thermal-hydraulic, and other applicable design-specific characteristics that either replicate the most recent fuel load in the reference commercial nuclear plant or, prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the removal of the features to prevent criticality required under \u00a7 53.620(d)(1)), replicate the intended initial fuel load for the reference commercial nuclear plant, with the exception of those portions of the simulation facility that utilize the reference plant itself.\n\n(iii) Simulation facility fidelity must be demonstrated so that significant control manipulations are completed without procedural exceptions, simulator performance exceptions, or deviation from the approved training scenario sequence.\n\n(3) Facility licensees that maintain a simulation facility that has been approved by the Commission for training purposes, demonstrating compliance with experience requirements, or the conduct of examinations under \u00a7 53.780(b) and (c) for the facility licensee's reference plant must:\n\n(i) Conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraph (e)(2) of this section is met;\n\n(ii) Retain the results of performance testing for 4 years after the completion of each performance test or until superseded by updated test results;\n\n(iii) Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification as to why the presence of such discrepancies will not adversely affect simulator performance with respect to the criteria of paragraph (e)(2) of this section;\n\n(iv) Make the results of any uncorrected performance test failures that may exist at the time of the initial license examination or requalification examination available for NRC review, prior to or concurrent with preparations for each initial license examination or requalification examination; and\n\n(v) Maintain the provisions for license application and examination integrity consistent with \u00a7 53.780(d).\n\n(4) A simulation facility must demonstrate compliance with the requirements of paragraphs (e)(2) and (e)(3) of this section for the Commission to accept the simulation facility for conducting initial examinations as described in \u00a7 53.780(b), requalification training as described in \u00a7 53.780(c), or performing control manipulations that affect reactivity to establish eligibility for an operator or senior operator license as described in \u00a7 53.775(a).\n\n(f)  Waiver of examination requirement.  On application, the Commission may waive any or all of the requirements for an initial licensing examination if it finds that the applicant has demonstrated the required knowledge, skills, and abilities to safely operate the plant, and is capable of continuing to do so. The Commission may make such a finding based on demonstration of the following:\n\n(1) Recent operating experience at a comparable facility;\n\n(2) Proof of the applicant's past competent and safe performance; and\n\n(3) Proof of the applicant's current qualifications.\n\n(g)  Proficiency.  The facility licensee must develop, implement, and maintain a proficiency program to ensure that operators and senior operators will actively perform the functions of an operator or senior operator, respectively, as needed to maintain proficiency with on-shift duties and familiarity with plant status. This program must include those steps that will be taken to re-establish proficiency when it cannot be maintained. This program must be approved by the Commission as part of its approval of the OL or COL for the plant. The approved proficiency program is subject to the requirements of \u00a7 53.1565.\n\n(h)  Records.  Each record required by this section must be legible throughout the retention period specified by each Commission regulation. The record may be the original, a reproduced copy, or an electronic copy provided that the copy is authenticated by authorized personnel."], ["10:10:2.0.1.1.3.6.62.17", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.785 Conditions of operator and senior operator licenses.", "NRC", "", "", "", "Each operator and senior operator license contains and is subject to the following conditions whether stated in the license or not:\n\n(a) Neither the license nor any right under the license may be assigned or otherwise transferred.\n\n(b) The license is limited to the facility or facilities for which it is issued.\n\n(c) The license is limited to those controls of the facility or facilities specified in the license.\n\n(d) The license is subject to, and the licensee must observe, all applicable rules, regulations, and orders of the Commission.\n\n(e) The licensee must maintain or re-establish proficiency in accordance with the facility licensee's Commission-approved proficiency program required under \u00a7 53.780(g).\n\n(f) The licensee must be subject to the facility's Commission-approved operator licensing requalification and requalification examination programs required under \u00a7 53.780(c).\n\n(g) The licensee must have a biennial medical examination as described by \u00a7 53.765.\n\n(h) The licensee must notify the Commission within 30 days about a conviction for a felony.\n\n(i) The licensee must not consume or ingest alcoholic beverages within the protected area of commercial nuclear plants. The licensee must not use, possess, or sell any illegal drugs. The licensee must not perform activities authorized by a license issued under this part while under the influence of alcohol or any prescription, over-the-counter, or illegal substance that could adversely affect his or her ability to safely and competently perform his or her licensed duties. For the purpose of this paragraph (i), with respect to alcoholic beverages and drugs, the term \u201cunder the influence\u201d means the licensee exceeded, as evidenced by a confirmed test result, the lower of the cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as established by the facility licensee. The term \u201cunder the influence\u201d also means the licensee could be mentally or physically impaired as a result of substance use including prescription and over-the-counter drugs, as determined under the provisions, policies, and procedures established by the facility licensee for its fitness-for-duty program, in such a manner as to adversely affect his or her ability to safely and competently perform licensed duties.\n\n(j) Each licensee must participate in the drug and alcohol testing programs as required under 10 CFR part 26.\n\n(k) The licensee must comply with any other conditions that the Commission may impose to protect health or to minimize danger to life or property."], ["10:10:2.0.1.1.3.6.62.18", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.790 Issuance, modification, and revocation of operator and senior operator licenses.", "NRC", "", "", "", "(a)  Issuance of operator and senior operator licenses.  If the Commission determines that an applicant for an operator license or a senior operator license demonstrates compliance with the requirements of the Atomic Energy Act of 1954, as amended, (the Act) and its regulations, it will issue a license in the form and containing any conditions and limitations it considers appropriate and necessary.\n\n(b)  Modification and revocation of operator and senior operator licenses.  (1) The terms and conditions of all operator and senior operator licenses are subject to amendment, revision, or modification by reason of rules, regulations, or orders issued in accordance with the Act or any amendments thereto.\n\n(2) Any license may be revoked, suspended, or modified, in whole or in part\u2014\n\n(i) For any material false statement in the application or in any statement of fact required under section 182 of the Act;\n\n(ii) Because of conditions revealed by the application or statement of fact or any report, record, inspection, or other means that would warrant the Commission to refuse to grant a license on an original application;\n\n(iii) For willful violation of, or failure to observe, any of the terms and conditions of the Act or the license, or of any rule, regulation, or order of the Commission;\n\n(iv) For any conduct determined by the Commission to be a hazard to safe operation of the facility; or\n\n(v) For the sale, use, or possession of illegal drugs, or refusal to participate in the facility drug and alcohol testing program, or a confirmed positive test for drugs, drug metabolites, or alcohol in violation of the conditions and cutoff levels established by \u00a7 53.785(i) or the consumption of alcoholic beverages within the protected area of commercial nuclear plants, or a determination of unfitness for scheduled work as a result of the consumption of alcoholic beverages."], ["10:10:2.0.1.1.3.6.62.19", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.795 Expiration and renewal of operator and senior operator licenses.", "NRC", "", "", "", "(a)  Expiration.  (1) Each operator license and senior operator license expires 6 years after the date of issuance, upon termination of employment with the facility licensee, or upon determination by the facility licensee that the licensed individual no longer needs to maintain a license.\n\n(2) If a licensee files an application for renewal or an upgrade of an existing license on NRC Form 398 at least 30 days before the expiration of the existing license, it does not expire until disposition of the application for renewal or for an upgraded license has been finally determined by the Commission. Filing by mail will be deemed to be complete at the time the application is postmarked\n\n(b)  Renewal.  (1) The applicant for renewal of an operator license or senior operator license must\u2014\n\n(i) Complete and sign NRC Form 398 and include the number of the license for which renewal is sought.\n\n(ii) File an original of NRC Form 398 as specified in \u00a7 53.775.\n\n(iii) Provide written evidence of the applicant's experience under the existing license and the approximate number of hours that the licensee has operated the facility.\n\n(iv) Provide a statement by an authorized representative of the facility licensee that during the effective term of the current license the applicant has satisfactorily completed the requalification program for the facility for which operator or senior operator license renewal is sought.\n\n(v) Provide evidence that the applicant has discharged the license responsibilities competently and safely. The Commission may accept as evidence of the applicant's having met this requirement a certificate of an authorized representative of the facility licensee or holder of an authorization by which the licensee has been employed.\n\n(vi) Provide certification by the facility licensee of medical condition and general health on NRC Form 396, to comply with \u00a7 53.765.\n\n(2) The license will be renewed if the Commission finds that\u2014\n\n(i) The medical condition and the general health of the licensee continue to be such as not to cause operational errors that endanger public health and safety. The Commission will base this finding upon the certification by the facility licensee as described in \u00a7 53.765(b).\n\n(ii) The licensee\u2014\n\n(A) Is capable of continuing to competently and safely assume licensed duties;\n\n(B) Has successfully completed a requalification program that has been approved by the Commission as required by \u00a7 53.780(c); and\n\n(C) Has passed the requalification examinations as required by \u00a7 53.780(c).\n\n(iii) There is a continued need for an operator to operate or for a senior operator to supervise operators at the facility designated in the application.\n\n(iv) The past performance of the licensee has been satisfactory to the Commission. In making its finding, the Commission will include in its evaluation information such as notices of violations or letters of reprimand in the licensee's docket."], ["10:10:2.0.1.1.3.6.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.710 Maintaining capabilities and availability of structures, systems, and components.", "NRC", "", "", "", "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 \u00a7\u00a7 53.210 and 53.220 will be met.\n\n(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 \u00a7 53.230 and support meeting the safety criteria of \u00a7 53.210. The technical specifications must describe the following requirements:\n\n(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 \u00a7 53.210 as a result of a design-basis accident analyzed in accordance with \u00a7 53.450(f).\n\n(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 \u00a7 53.210.\n\n(3) For each SSC classified as SR in accordance with \u00a7 53.460, technical specifications must define\u2014\n\n(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 \u00a7 53.450(f) satisfy the safety criteria of \u00a7 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.\n\n(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.\n\n(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.\n\n(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 \u00a7 53.040.\n\n(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 \u00a7\u00a7 53.220 and 53.450(e).\n 1 \n   The control measures must\u2014\n\n1  The comprehensive risk metrics and related risk performance objectives established under \u00a7 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.\n\n(1)(i) Identify who within the licensee's organization has authority to make configuration changes;\n\n(ii) Establish processes to make configuration changes to NSRSS SSCs; and\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.6.62.20", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.800 Facility licensees for self-reliant-mitigation facilities.", "NRC", "", "", "", "(a) A commercial nuclear plant is a self-reliant-mitigation facility if the NRC determined as part of its approval of the OL or COL for that plant that its design demonstrates compliance with the criteria in paragraphs (a)(1) though (a)(5) of this section. A self-reliant-mitigation facility is of a class, based upon the similarity of operating and technical characteristics of the plants in the class, such that its licensee must comply with the requirements of \u00a7\u00a7 53.800 through 53.820 in lieu of those in \u00a7\u00a7 53.760 through 53.795.\n\n(1) The safety performance criteria of \u00a7\u00a7 53.210 and 53.220 must be met without reliance upon human action for credited event mitigation.\n\n(2) The results of the probabilistic risk assessment (PRA), other systematic risk evaluations, or a combination thereof required by \u00a7 53.450(a) must demonstrate that the evaluation criteria for the events analyzed in accordance with \u00a7 53.450 will be met without reliance on human actions to achieve acceptable event mitigation.\n\n(3) The functional requirements analysis and function allocation performed under \u00a7 53.730(d) must demonstrate that functions required for safety are not reliant upon credited human action.\n\n(4) The plant response to events analyzed under \u00a7 53.450 must rely exclusively on safety features and characteristics that will neither be rendered unavailable by credible human errors of commission or omission nor credibly require manual human operation in response to equipment failures. Compliance with this paragraph (a)(4) may be achieved through the use of SSCs that function through inherent characteristics or that have engineered protections against human failures.\n\n(5) Assessments of credited human actions within the analysis of design-basis accidents (DBAs) and across the range of LBEs other than DBAs do not identify important human actions needed to ensure appropriate defense in depth is provided, as required by \u00a7 53.250.\n\n(b) [Reserved]"], ["10:10:2.0.1.1.3.6.62.21", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.805 Facility licensee requirements related to generally licensed reactor operators.", "NRC", "", "", "", "(a) Licensees 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 \u00a7 53.1070 must demonstrate compliance with the following requirements:\n\n(1) Ensure that, in addition to being qualified to perform those items identified by the facility-specific systems approach to training conducted under \u00a7 53.815, generally licensed reactor operators are qualified to safely and competently\u2014\n\n(i) Perform administrative tasks, including compliance with technical specifications, and perform operability determinations;\n\n(ii) Implement maintenance and configuration controls;\n\n(iii) Comply with radioactive release limitations;\n\n(iv) Understand plant operating data, including reactor parameters, and evaluate emergency conditions;\n\n(v) Initiate a reactor shutdown from necessary locations;\n\n(vi) Dispatch and direct operations and maintenance personnel;\n\n(vii) Implement any applicable responsibilities under the facility emergency plan; and\n\n(viii) Make required notifications to local, State, participating Tribal, and Federal authorities.\n\n(2) Develop, implement, and maintain facility technical specifications that provide the necessary administrative controls to ensure the implementation of the requirements in this section.\n\n(3) Develop, implement, and maintain the generally licensed reactor operator training, examination, and proficiency programs required under \u00a7 53.815.\n\n(4) Ensure that generally licensed reactor operators are subject to the facility's generally licensed reactor operator training, examination, and proficiency programs required under \u00a7 53.815. Ensure that generally licensed reactor operators are subject to and comply with the applicable programmatic requirements for personnel required under 10 CFR parts 26 and 73. An individual that is not in compliance with any of these programs is not qualified to be in a position that may involve the manipulation of the controls of the commercial nuclear plant.\n\n(5) Report annually to the NRC the identity of all generally licensed reactor operators at the commercial nuclear plant, including all additions and deletions since the previous report.\n\n(6) Ensure that the facility design continues to meet the criteria of \u00a7 53.800.\n\n(b) [Reserved]"], ["10:10:2.0.1.1.3.6.62.22", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.810 Generally licensed reactor operators.", "NRC", "", "", "", "(a) A general license to manipulate the controls of a self-reliant-mitigation facility and to direct the licensed activities of generally licensed reactor operators is hereby issued to any individual employed in a position that may involve the manipulation of the controls of that self-reliant-mitigation facility and who observes the restrictions of this section.\n\n(b) A generally licensed reactor operator must comply with the operating procedures and other conditions specified in the license authorizing operation of the facility.\n\n(c) The general license is limited to the facility or facilities at which the operator is employed.\n\n(d) The Commission will suspend the general license on an individual operator basis for violations of any provision of the Act or any rule or regulation issued thereunder whenever the Commission deems such suspension desirable, including\u2014\n\n(1) For willful violation of, or failure to observe, any of the terms and conditions of the Act or the general license, or of any rule, regulation, or order of the Commission;\n\n(2) For any conduct determined by the Commission to be a hazard to safe operation of the facility; or\n\n(3) For the sale, use, or possession of illegal drugs, or refusal to participate in the facility drug and alcohol testing program, or a confirmed positive test for drugs, drug metabolites, or alcohol in violation of the conditions and cutoff levels established by \u00a7 53.810(f) or the consumption of alcoholic beverages within the protected area of commercial nuclear plants, or a determination of unfitness for scheduled work as a result of the consumption of alcoholic beverages.\n\n(e) The Commission may require information from a generally licensed reactor operator to determine whether a general license should be revoked or suspended with respect to that operator.\n\n(f) The generally licensed reactor operator must not consume or ingest alcoholic beverages within the protected area of commercial nuclear plants. The generally licensed reactor operator must not use, possess, or sell any illegal drugs. The generally licensed reactor operator must not perform activities requiring a general license while under the influence of alcohol or any prescription, over-the-counter, or illegal substance that could adversely affect his or her ability to safely and competently perform these activities. For the purpose of this paragraph (f), with respect to alcoholic beverages and drugs, the term \u201cunder the influence\u201d means the generally licensed reactor operator exceeded, as evidenced by a confirmed test result, the lower of the cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as established by the facility licensee. The term \u201cunder the influence\u201d also means the generally licensed reactor operator could be mentally or physically impaired as a result of substance use including prescription and over-the-counter drugs, as determined under the provisions, policies, and procedures established by the facility licensee for its fitness-for-duty program, in such a manner as to adversely affect his or her ability to safely and competently perform generally licensed reactor operator duties.\n\n(g) The generally licensed reactor operator must notify the Commission within 30 days about a conviction for a felony."], ["10:10:2.0.1.1.3.6.62.23", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.815 Generally licensed reactor operator training, examination, and proficiency programs.", "NRC", "", "", "", "(a)  Applicability.  The requirements of this section apply to each licensee of a self-reliant-mitigation facility that has not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under \u00a7 53.1070.\n\n(b)  Requirements.  (1) The facility licensee must develop, implement, and maintain training and examination programs that demonstrate compliance with the requirements of paragraphs (b)(2) through (b)(3) of this section.\n\n(2) The training program must provide for both the initial and continuing training of generally licensed reactor operators and be derived from a systems approach to training as defined in this part.\n\n(3)(i) The training program must incorporate the instructional requirements necessary to provide qualified generally licensed reactor operators to operate and maintain the facility in a safe manner in all modes of operation. The training program must comply with the facility license, including all technical specifications and applicable regulations. The facility licensee must periodically evaluate and revise the training program as appropriate to reflect industry experience and relevant changes, including changes to the facility, procedures, regulations, and quality assurance (QA) requirements. Facility licensee management must periodically review the training program for effectiveness.\n\n(ii) The training program must ensure that generally licensed reactor operators have and maintain the necessary knowledge, skills, and abilities.\n\n(iii) The training program must include the generally licensed reactor operator manipulating the controls of either the facility or a simulation facility that demonstrates compliance with the requirements of \u00a7 53.815(e).\n\n(iv) The training program must include an initial examination program for testing a representative sample of the knowledge, skills, and abilities needed to safely perform generally licensed reactor operator duties, to include both the examination methods and criteria to be used to assess passing performance. The facility licensee must provide the opportunity for a representative of the Commission to be present during initial examination administration.\n\n(v) The training program must include a requalification examination program for testing a sample of the topics included under the systems approach to training, to include the examination methods and criteria to be used to assess passing performance. The requalification examination program must specify an appropriate periodicity for administering a complete requalification examination to each generally licensed reactor operator, and the facility licensee must provide the opportunity for a representative of the Commission to be present during requalification examination administration.\n\n(A) The facility licensee must ensure that any generally licensed reactor operator who either demonstrates unsatisfactory performance on, or fails to complete, the requalification examination is removed from the performance of generally licensed reactor operator duties until such time that any necessary remedial training has been completed and a retake examination has been passed.\n\n(B) [Reserved]\n\n(vi) The training program must be approved by the Commission prior to its use, as described under \u00a7 53.730(g). The examination program must provide for valid and reliable examinations and must be approved by the Commission prior to their use, as described under \u00a7 53.730(g). The approved programs are subject to the requirements of \u00a7 53.1565.\n\n(c)  Records.  The following is required regarding the documentation of the generally licensed reactor operator training and examination programs:\n\n(1) Sufficient records must be maintained by the facility licensee to maintain the integrity of the programs and kept available for NRC inspection to verify the adequacy of the programs.\n\n(2) The facility licensee must maintain records documenting the participation of each generally licensed reactor operator in the training and examination programs. The records must contain copies of examinations administered, the answers given by the generally licensed reactor operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which a generally licensed reactor operator exhibited deficiencies. The facility licensee must retain these records while the associated generally licensed reactor operators remain employed at the facility.\n\n(3) Each record required by this part must be legible throughout the retention period. The record may be the original, a reproduced copy, or an electronic copy provided that the copy is authenticated by authorized personnel.\n\n(d)  Examination integrity.  Generally licensed reactor operators and facility licensees must not engage in any activity that compromises the integrity of any examination conducted under the generally licensed reactor operator training and examination programs. The integrity of an examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, could have affected the consistent administration of the examination. This includes all activities related to the preparation, administration, and grading of examinations.\n\n(e)  Simulation facilities.  (1) Simulation facilities used for training purposes, for maintaining proficiency, or for the conduct of examinations must demonstrate compliance with the following criteria as they relate to the facility licensee's reference plant:\n\n(i) The simulation facility must be of sufficient scope and fidelity for individuals to acquire and demonstrate the necessary knowledge, skills, and abilities to safely perform generally licensed reactor operator duties.\n\n(ii) The simulation facility must utilize models relating to nuclear, thermal-hydraulic, and other applicable design-specific characteristics that either replicate the most recent fuel load in the reference commercial nuclear plant or, prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the removal of the features to prevent criticality required under \u00a7 53.620(d)(1)), replicate the intended initial fuel load for the reference commercial nuclear plant, with the exception of those portions of the simulation facility that utilize the reference plant itself.\n\n(iii) Simulator fidelity must be demonstrated so that significant control manipulations are completed without procedural exceptions, simulator performance exceptions, or deviation from the approved training scenario sequence.\n\n(2) Facility licensees that maintain a simulation facility for training purposes, for maintaining proficiency, or for the conduct of examinations must\u2014\n\n(i) Conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraph (e)(1) of this section is met;\n\n(ii) Retain the results of performance testing for 4 years after the completion of each performance test or until superseded by updated test results;\n\n(iii) Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification for why the presence of such discrepancies will not adversely affect the criteria of paragraph (e)(1) of this section;\n\n(iv) Make the results of any uncorrected performance test failures that may exist at the time of an inspection available for NRC review; and\n\n(v) Maintain the provisions for examination integrity consistent with \u00a7 53.815(d).\n\n(f)  Waiver of examination requirement.  The facility licensee may waive any or all of the requirements for an examination in accordance with the facility licensee's Commission-approved generally licensed reactor operator training and examination programs.\n\n(g)  Proficiency.  The facility licensee must develop, implement, and maintain a proficiency program to allow generally licensed reactor operators to maintain proficiency regarding position functions and familiarity with plant status. This program must include those steps that will be taken in order to re-establish proficiency when it cannot be maintained."], ["10:10:2.0.1.1.3.6.62.24", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.820 Cessation of individual applicability.", "NRC", "", "", "", "The general license ceases to be applicable on an individual basis once a generally licensed reactor operator is no longer being employed in a position that may involve the manipulation of the controls of the self-reliant mitigation facility."], ["10:10:2.0.1.1.3.6.62.25", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.830 Training and qualification of commercial nuclear personnel.", "NRC", "", "", "", "(a) This section addresses personnel training requirements. The regulations within this section are applicable to all applicants for or holders of OLs or COLs under this part.\n\n(b) Prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the removal of the features to prevent criticality required under \u00a7 53.620(d)(1)), each holder of an operating or COL under this part must, with sufficient time to provide trained and qualified personnel to operate the facility, establish, implement, and maintain a training program that demonstrates compliance with the requirements of paragraphs (c) and (d) of this section.\n\n(c) The training program must be derived from a systems approach to training as defined in this part and must provide, at a minimum, for the training and qualification of the following categories of commercial nuclear personnel:\n\n(1) Supervisors ( e.g.,  shift supervisors);\n\n(2) Technicians ( e.g.,  maintenance, chemistry, and radiological); and\n\n(3) Other appropriate operating personnel ( e.g.,  auxiliary operators, certified fuel handlers, and individuals who provide engineering expertise to on-shift operating personnel).\n\n(d) The training program must incorporate the instructional requirements necessary to provide qualified personnel to operate components of a commercial nuclear plant and maintain the facility in a safe manner in all modes of operation. The training program must be developed to be in compliance with the facility license, including all technical specifications and applicable regulations.\n\n(1) The training program must be periodically evaluated and revised as appropriate to reflect industry experience and relevant changes, including changes to the facility, procedures, regulations, and QA requirements. The training program must be periodically reviewed by facility licensee management for effectiveness.\n\n(2) Sufficient records must be maintained by the facility licensee to maintain program integrity and kept available for NRC inspection to verify the adequacy of the training program."], ["10:10:2.0.1.1.3.6.62.26", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.845 Programs.", "NRC", "", "", "", "(a) The required plant programs under this part must include but are not necessarily limited to the programs described in the following sections of this subpart. Licensees may combine, separate, and otherwise organize programs and related documents as appropriate for the technologies and organizations associated with the commercial nuclear plant.\n\n(b) In addition to the programs described in the following sections, programs must be provided for each commercial nuclear plant, if necessary, to ensure that the performance of design features and human actions are consistent with the analyses performed under \u00a7\u00a7 53.450 and 53.730 and that the plant will demonstrate compliance with the safety criteria defined in \u00a7\u00a7 53.210 and 53.220."], ["10:10:2.0.1.1.3.6.62.27", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.850 Radiation protection.", "NRC", "", "", "", "(a) Each holder of an OL or COL under this part must develop, implement, and maintain a Radiation Protection Program for operations that is commensurate with the scope and extent of licensed activities under this part and includes measures for limiting and monitoring radioactive plant effluents and limiting and monitoring the dose to individuals working with radioactive materials in accordance with 10 CFR part 20.\n\n(b) Each holder of an OL or COL under this part must develop, implement, and maintain a program for the control of radioactive effluents and for environmental monitoring. The program must be contained in an Offsite Dose Calculations Manual, must be implemented by procedures, and must include remedial actions to be taken whenever the program limits are exceeded. The Offsite Dose Calculations Manual must\u2014\n\n(1) Contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and\n\n(2) Contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by \u00a7 53.1645.\n\n(c) Each holder of an OL or COL under this part must develop, implement, and maintain a Process Control Program that identifies the administrative and operational controls for solid radioactive waste processing, process parameters, and surveillance requirements sufficient to ensure compliance with the requirements of 10 CFR part 20, 10 CFR part 61, and 10 CFR part 71."], ["10:10:2.0.1.1.3.6.62.28", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.855 Emergency preparedness.", "NRC", "", "", "", "(a) Each holder of an OL or COL under this part must have an emergency response plan that must contain information needed to demonstrate compliance with either the requirements in \u00a7 50.160 of this chapter or the requirements in appendix E to part 50 and the planning standards of \u00a7 50.47(b) of this chapter.\n\n(b) No initial OL, initial COL, or early site permit that includes complete and integrated emergency plans will be issued under this part unless a finding is made by the NRC, in accordance with \u00a7 50.47 of this chapter, that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."], ["10:10:2.0.1.1.3.6.62.29", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.860 Security programs.", "NRC", "", "", "", "(a)  Physical protection program.  Each holder of an OL or COL under this part must develop, implement, and maintain a physical protection program under the following requirements:\n\n(1) The licensee must implement security requirements for the protection of special nuclear material based on the type, enrichment, and quantity in accordance with 10 CFR part 73, as applicable, and implement security requirements for the protection of Category 1 and Category 2 quantities of radioactive material in accordance with 10 CFR part 37, as applicable; and\n\n(2) The licensee must demonstrate compliance with the provisions set forth in either \u00a7 73.55 or \u00a7 73.100 of this chapter.\n\n(b)  Fitness-for-duty.  Each holder of an OL or COL under this part must develop, implement, and maintain a fitness-for-duty program under 10 CFR part 26.\n\n(c)  Access authorization.  Each holder of an OL or COL under this part must develop, implement, and maintain an access authorization program under \u00a7 73.56 or \u00a7 73.120 of this chapter, as applicable.\n\n(d)  Cybersecurity.  Each holder of an OL or COL under this part must develop, implement, and maintain a cybersecurity program under \u00a7 73.54 or \u00a7 73.110 of this chapter.\n\n(e)  Information security.  Each holder of an OL or COL under this part must develop, implement, and maintain an information protection system under \u00a7\u00a7 73.21, 73.22, and 73.23 of this chapter, as applicable."], ["10:10:2.0.1.1.3.6.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.715 Maintenance, repair, and inspection programs.", "NRC", "", "", "", "(a) A program to control maintenance activities and monitor the performance or condition of SR and NSRSS SSCs must be developed, implemented, and maintained.\n\n(b) Whenever a licensee determines through activities related to maintenance, repair, and inspection of SSCs, the activities under \u00a7 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.\n\n(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.\n\n(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."], ["10:10:2.0.1.1.3.6.62.30", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.865 Quality assurance.", "NRC", "", "", "", "Each holder of an OL or COL under this part must develop, implement, and maintain a quality assurance program in accordance with appendix B of part 50 of this chapter. A written quality assurance program manual must be developed and used to guide the conduct of the program."], ["10:10:2.0.1.1.3.6.62.31", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.870 Integrity assessment programs.", "NRC", "", "", "", "Each holder of an OL or COL under this part must develop, implement, and maintain an integrity assessment program to monitor, evaluate, and manage\u2014\n\n(a) The effects of plant aging on SR and NSRSS SSCs. The program may refer to surveillances, tests, and inspections conducted for specific SSCs in accordance with other requirements in this part or conducted in accordance with applicable consensus codes and standards endorsed or otherwise found acceptable by the NRC;\n\n(b) Cyclic or transient load limits to ensure that SR and NSRSS SSCs are maintained within the applicable design limits; and\n\n(c) Degradation mechanisms related to chemical interactions, operating temperatures, effects of irradiation, and other environmental factors to ensure that the capabilities, availability, and reliability of SR and NSRSS SSCs demonstrate compliance with the functional design criteria of \u00a7\u00a7 53.410 and 53.420."], ["10:10:2.0.1.1.3.6.62.32", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.875 Fire protection.", "NRC", "", "", "", "(a)(1) Each holder of an OL or COL under this part must have a fire protection plan that describes the overall fire protection program for the facility; identifies the various positions within the licensee's organization that are responsible for the program; states the authorities that are delegated to each of these positions to implement those responsibilities; and outlines the plans for fire protection, fire detection and suppression capability; and limitation of fire damage.\n\n(2) The fire protection plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as the following: administrative controls and personnel requirements for fire prevention and manual fire suppression activities; automatic and manually operated fire detection and suppression systems; and the means to limit fire damage to SSCs so that the capability to demonstrate compliance with the requirements of \u00a7 53.210 is ensured.\n\n(b)(1) Each holder of an OL or COL under this part must develop a performance-based or deterministic fire protection program that demonstrates compliance with the safety criteria outlined in \u00a7\u00a7 53.210 and 53.220, related safety functions outlined in \u00a7 53.230, and defense in depth as outlined in \u00a7 53.250 with specific fire protection measures related to fire prevention, fire detection, and fire suppression.\n\n(2) The fire protection program must comply with the following:\n\n(i) Safety-related and, where appropriate, NSRSS SSCs must be designed, located, and maintained to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.\n\n(ii) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs.\n\n(iii) Fire detection and fire suppression systems of appropriate capacity and capability must be provided and designed and maintained to minimize the adverse effects of fires on SR and NSRSS SSCs.\n\n(iv) Fire suppression systems must be designed and maintained 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 \u00a7 53.230."], ["10:10:2.0.1.1.3.6.62.33", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.880 Inservice inspection and inservice testing.", "NRC", "", "", "", "(a) Each holder of an OL or COL under this part must develop, implement, and maintain a program for inservice inspection (ISI) and inservice testing (IST) prior to receiving an OL or COL. The ISI/IST programs must, wherever applicable, be in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC. The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be supplemented by risk insights that identify the most important SSCs to plant safety. The types of testing and inspections and their frequency should be informed by risk insights to maintain the reliability and performance of SSCs consistent with the associated design and analyses activities involving those SSCs. Risk insights must also be used to determine when to conduct the inspections and tests ( e.g.,  full power, shutdown, refueling) to minimize risk to the plant workers and the public. The ISI/IST program must be documented in a written manual and managed by qualified personnel reporting to the director, responsible officer, or designated person.\n\n(b) Prior to plant operation, baseline inspections and testing must be performed using the same techniques as will be used for future inspections and testing. The results of these inspections and testing must be used as benchmarks for evaluating the results of future inspections and testing. Sufficient room and support must be provided to accommodate the personnel, ISI/IST equipment, and shielding necessary to perform the inspections and testing. Acceptance criteria for determining whether corrective action is needed must be developed (or taken from the codes and standards used in the design) for evaluating the results of the inspections and testing. The results of the inspections and testing must be provided to the director, responsible officer, or designated person who is responsible for determining what, if any, corrective action is needed and when it should be taken. The ISI/IST results and corrective actions must be documented and the documentation retained for the life of the plant."], ["10:10:2.0.1.1.3.6.62.34", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.910 Procedures and guidelines.", "NRC", "", "", "", "(a) Each holder of an OL or COL under this part must have a program for developing, implementing, and maintaining an integrated set of procedures, guidelines, and related supporting activities to support normal operations and respond to possible unplanned events.\n\n(b) The program required by paragraph (a) of this section must include but is not limited to development, implementation, maintenance, and supporting activities of procedures and guidelines for the following:\n\n(1) Plant operations;\n\n(2) Maintenance activities under \u00a7 53.715;\n\n(3) Program requirements under this subpart;\n\n(4) Emergency operating procedures, if developed to address the role of human actions in responding to LBEs;\n\n(5) Accident management guidelines, if developed to address the role of human actions in responding to LBEs;\n\n(6) Procedures for each area in which licensed special nuclear material is handled, used, or stored to protect personnel upon the sounding of a criticality alarm required by \u00a7 53.440(m); and\n\n(7) Procedures that describe how the licensee will address the following areas if the licensee is notified of a potential aircraft threat:\n\n(i) Verification of the authenticity of threat notifications;\n\n(ii) Maintenance of continuous communication with threat notification sources;\n\n(iii) Contacting all onsite personnel and applicable offsite response organizations;\n\n(iv) Onsite actions necessary to enhance the capability of the facility to mitigate the consequences of an aircraft impact;\n\n(v) Measures to reduce visual discrimination of the site relative to its surroundings or individual buildings within the protected area;\n\n(vi) Dispersal of equipment and personnel, as well as rapid entry into site protected areas for essential onsite personnel and offsite responders who are necessary to mitigate the event; and\n\n(vii) Recall of site personnel."], ["10:10:2.0.1.1.3.6.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.720 Response to seismic events.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.6.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.725 General staffing, training, personnel qualifications, and human factors requirements.", "NRC", "", "", "", "(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 \u00a7 53.800(a)(1) through (a)(5). Otherwise, the commercial nuclear plant is an interaction-dependent-mitigation facility.\n\n(b)  Purpose and applicability.  The regulations in \u00a7\u00a7 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:\n\n(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.\n\n(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 \u00a7 53.1070.\n\n(3) Sections 53.800 through 53.820 address generally licensed reactor operator requirements. The regulations within these sections are in lieu of \u00a7\u00a7 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 \u00a7 53.1070.\n\n(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.\n\n(c)  Definitions.  When used in \u00a7\u00a7 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 \u00a7\u00a7 53.725 through 53.830\u2014\n\nAutomation  means a device or system that accomplishes (partially or fully) a function or task.\n\nAuxiliary 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.\n\nControls  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.\n\nGenerally licensed reactor operator  means any individual licensed under the provisions of \u00a7 53.810 to manipulate controls of a self-reliant-mitigation facility and to direct the licensed activities of generally licensed reactor operators.\n\nInteraction-dependent-mitigation facility  means a commercial nuclear plant design other than one that demonstrates compliance with the operating and technical characteristics defined under \u00a7 53.800.\n\nLoad following  means a commercial nuclear plant automatically changing its output to match expected demand in response to externally originated instructions or signals.\n\nOperator  means any individual licensed under the provisions of \u00a7\u00a7 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility.\n\nPerformance testing  means testing conducted to verify a simulation facility's performance as compared to actual or predicted reference plant performance.\n\nReference 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.\n\nSelf-reliant-mitigation facility  means a commercial nuclear plant design that demonstrates compliance with the operating and technical characteristics defined under \u00a7 53.800.\n\nSenior operator  means any individual licensed under the provisions of \u00a7\u00a7 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility and to direct the licensed activities of operators.\n\nSimulation 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.\n\nSystems approach to training  means a training program that includes the following five elements:\n\n(i) Systematic analysis of the jobs to be performed.\n\n(ii) Learning objectives derived from the analysis which describe desired performance after training.\n\n(iii) Training design and implementation based on the learning objectives.\n\n(iv) Evaluation of trainee mastery of the objectives during training.\n\n(v) Evaluation and revision of the training based on the performance of trained personnel in the job setting."], ["10:10:2.0.1.1.3.6.62.6", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.726 Communications.", "NRC", "", "", "", "(a) An applicant or licensee or facility licensee must submit any communication or report required by the regulations contained within \u00a7\u00a7 53.725 through 53.830 and must submit any application filed under these regulations to the Commission.\n\n(b) Each facility licensee that is required to comply with the requirements of \u00a7\u00a7 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:\n\n(1) Permanent reassignment from the position for which the facility licensee has certified the need for a licensed operator or senior operator under \u00a7 53.775(a)(1);\n\n(2) Termination of any operator or senior operator; or\n\n(3) Permanent disability or illness as required under \u00a7 53.770."], ["10:10:2.0.1.1.3.6.62.7", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.728 Completeness and accuracy of information.", "NRC", "", "", "", "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."], ["10:10:2.0.1.1.3.6.62.8", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.730 Defining, fulfilling, and maintaining the role of personnel in ensuring safe operations.", "NRC", "", "", "", "Each applicant for or holder of an OL or COL for a commercial nuclear plant under this part must comply with the following:\n\n(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.\n\n(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:\n\n(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;\n\n(2) Automatic indication of the bypassed and operable status of safety systems;\n\n(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;\n\n(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;\n\n(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;\n\n(6) Monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of normal operating and accident conditions; and\n\n(7) For self-reliant-mitigation facilities, the plant design must also provide the generally licensed reactor operators with the capability to do the following:\n\n(i) Receive plant operating data, including reactor parameters and information needed for the evaluation of emergency conditions.\n\n(ii) Promptly dispatch operations and maintenance personnel.\n\n(iii) Immediately implement responsibilities under the facility emergency plan, as applicable.\n\n(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.\n\n(c)  Concept of operations.  A concept of operations that is of sufficient scope and detail to address the following must be provided:\n\n(1) Plant goals;\n\n(2) The roles and responsibilities of operating personnel and automation (or any combination thereof) that are responsible for completing plant functions;\n\n(3) Staffing, qualifications, and training;\n\n(4) The management of normal operations;\n\n(5) The management of off-normal conditions and emergencies;\n\n(6) The management of maintenance and modifications; and\n\n(7) The management of tests, inspections, and surveillances.\n\n(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:\n\n(1) The functional requirements analysis must address how safety functions and functional safety criteria are satisfied; and\n\n(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.\n\n(e)  Operating experience.  A program, during construction and during operation, as applicable, for evaluating and applying operating experience must be developed, implemented, and maintained.\n\n(f)  Staffing plan.  A staffing plan must be developed and comply with the following:\n\n(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:\n\n(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\n\n(ii) A Professional Engineer's license from a U.S. State or territory.\n\n(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.\n\n(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.\n\n(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.\n\n(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 \u00a7 53.1565.\n\n(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:\n\n(1) For those applicants for or holders of OLs or COLs for interaction-dependent-mitigation facilities:\n\n(i) The operator licensing initial training program required under \u00a7 53.780(a);\n\n(ii) The operator licensing initial examination program required under \u00a7 53.780(b);\n\n(iii) The operator licensing requalification program required under \u00a7 53.780(c); and\n\n(iv) The operator proficiency program required under \u00a7 53.780(g).\n\n(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 \u00a7 53.815.\n\n(3) The operator licensing requalification programs required under \u00a7 53.780(c) or \u00a7 53.815(b) must be implemented upon commencing the administration of initial examinations under the operator licensing examination program required under \u00a7 53.780(b) or \u00a7 53.815(b), respectively."], ["10:10:2.0.1.1.3.6.62.9", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "F", "Subpart F\u2014Requirements for Operation", "", "\u00a7 53.735 General exemptions.", "NRC", "", "", "", "The regulations in \u00a7\u00a7 53.725 through 53.830 do not require a license for an individual who\u2014\n\n(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\n\n(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."], ["10:10:2.0.1.1.3.7.62.1", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1000 Scope and purpose.", "NRC", "", "", "", "This subpart defines the requirements related to decommissioning for applicants for, or holders of, an operating license (OL) or combined license (COL). The requirements related to maintaining financial assurance for decommissioning are in \u00a7\u00a7 53.1010 through 53.1060. The requirements for transitioning from operations to decommissioning and for the release of property and termination of the license are in \u00a7\u00a7 53.1070 through 53.1080."], ["10:10:2.0.1.1.3.7.62.10", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1075 Program requirements during decommissioning.", "NRC", "", "", "", "(a) Licensees that have submitted the certifications required under \u00a7 53.1070 must maintain a decommissioning fire protection program to address the potential for fires that could cause the release or spread of radioactive materials.\n\n(1) The objectives of the decommissioning fire protection program are to\n\n(i) Reasonably prevent these fires from occurring;\n\n(ii) Rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and\n\n(iii) Ensure that the risk of fire-induced radiological hazards to the public, environment, and plant personnel is minimized.\n\n(2) The licensee must assess the decommissioning fire protection program on a regular basis. The licensee must revise the decommissioning fire protection program documentation as appropriate throughout the various stages of facility decommissioning.\n\n(3) The licensee may make changes to the decommissioning fire protection program without NRC approval if these changes do not reduce the effectiveness of fire protection for structures, systems, and components that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities.\n\n(b) [Reserved]"], ["10:10:2.0.1.1.3.7.62.11", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1080 Release of part of a commercial nuclear plant or site for unrestricted use.", "NRC", "", "", "", "(a) Prior written NRC approval is required to release part of a commercial nuclear plant or site for unrestricted use at any time before receiving approval of a license termination plan. Section 53.1060 specifies recordkeeping requirements associated with partial release. Holders of an OL or COL under this part seeking NRC review and approval must\u2014\n\n(1) Evaluate the effect of releasing the property to ensure that\u2014\n\n(i) The dose to individual members of the public does not exceed the limits and standards of subpart D of 10 CFR part 20;\n\n(ii) There is no reduction in the effectiveness of emergency planning or physical security;\n\n(iii) Effluent releases remain within license conditions;\n\n(iv) The environmental monitoring program and offsite dose calculation manual are revised to account for the changes;\n\n(v) The siting criteria of subpart D of this part continue to be met; and\n\n(vi) All other applicable statutory and regulatory requirements continue to be met.\n\n(2) Perform a historical site assessment of the part of the commercial nuclear plant or site to be released; and\n\n(3) Perform surveys adequate to demonstrate compliance with the radiological criteria for unrestricted use specified in \u00a7 20.1402 of this chapter for impacted areas.\n\n(b) For release of non-impacted areas, the licensee may submit a written request for NRC review and approval of the release if a license amendment is not otherwise required. The request submittal must include\u2014\n\n(1) The results of the evaluations performed in accordance with paragraphs (a)(1) and (a)(2) of this section;\n\n(2) A description of the part of the commercial nuclear plant or site to be released;\n\n(3) The schedule for release of the property;\n\n(4) The results of the evaluations performed in accordance with \u00a7 53.1540; and\n\n(5) A discussion that provides the reasons for concluding that the environmental impacts associated with the licensee's proposed release of the property will be bounded by appropriate previously issued environmental impact statements.\n\n(c) After receiving a request from the licensee for NRC approval of the release of a non-impacted area, the NRC must\u2014\n\n(1) Determine whether the licensee has adequately evaluated the effect of releasing the property as required by paragraph (a)(1) of this section;\n\n(2) Determine whether the licensee's classification of any release areas as non-impacted is adequately justified; and\n\n(3) If determining that the licensee's submittal is adequate, inform the licensee in writing that the release is approved.\n\n(d) For release of impacted areas, the licensee must submit an application for amendment of its license for the release of the property. The application must include\u2014\n\n(1) The information specified in paragraphs (b)(1) through (b)(3) of this section;\n\n(2) The methods used for and results obtained from the radiation surveys required to demonstrate compliance with the radiological criteria for unrestricted use specified in \u00a7 20.1402; and\n\n(3) A supplement to the environmental report, under \u00a7 51.53 of this chapter.\n\n(e) After receiving a license amendment application from the licensee for the release of an impacted area, the NRC must\u2014\n\n(1) Determine whether the licensee has adequately evaluated the effect of releasing the property as required by paragraph (a)(1) of this section;\n\n(2) Determine whether the licensee's classification of any release areas as non-impacted is adequately justified;\n\n(3) Determine whether the licensee's radiation survey for an impacted area is adequate; and\n\n(4) If determining that the licensee's submittal is adequate, approve the licensee's amendment application.\n\n(f) The NRC must publish notice receipt of the release approval request or license amendment application in the  Federal Register  and make the approval request or license amendment application available for public comment. Before acting on an approval request or license amendment application submitted in accordance with this section, the NRC must conduct a public meeting readily accessible to individuals in the vicinity of the licensee's facility for the purpose of obtaining public comments on the proposed release of part of the commercial nuclear plant or site. The NRC must publish a document in the  Federal Register  and in a forum, such as local newspapers, which is readily accessible to individuals in the vicinity of the site, announcing the date, time, and location of the meeting, along with a brief description of the purpose of the meeting."], ["10:10:2.0.1.1.3.7.62.2", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1010 Financial assurance for decommissioning.", "NRC", "", "", "", "(a) This section establishes requirements for indicating to the U.S. Nuclear Regulatory Commission (NRC) how an applicant for or holder of an OL or COL under this part will provide reasonable assurance that funds will be available for the decommissioning process. Reasonable assurance consists of a series of steps as provided in paragraph (b) of this section and \u00a7\u00a7 53.1020, 53.1030 and 53.1040. Funding for the decommissioning of commercial nuclear plants may also be subject to the regulation of Federal or State government agencies ( e.g.,  Federal Energy Regulatory Commission (FERC) and State Public Utility Commissions) that have jurisdiction over rate regulation. The requirements of this subpart, in particular \u00a7 53.1020, are in addition to, and not a substitution for, other requirements, and are not intended to be used by themselves or by other agencies to establish rates.\n\n(b) Each applicant for an OL or COL under this part must prepare a plan and an associated decommissioning report that ensures and documents that adequate funding will be available to decommission the facility. Each holder of an OL or COL must implement and maintain the plan.\n\n(1)(i) Before the Commission issues an OL under this part, the applicant must update the decommissioning report to certify that it has provided financial assurance for decommissioning in the amount proposed in the application and approved by the NRC under \u00a7 53.1020.\n\n(ii) No later than 30 days after the Commission issues the notice of intended operation under \u00a7 53.1452 for a COL under this part, the licensee must update the decommissioning report to certify that it has provided financial assurance for decommissioning in the amount proposed in the application and approved by the NRC under \u00a7 53.1020.\n\n(2) The amount of financial assurance for decommissioning to be provided must be based on a site-specific cost estimate for decommissioning the facility under \u00a7 53.1020.\n\n(3) The amount of financial assurance for decommissioning to be provided must be adjusted annually using a rate at least equal to that stated in \u00a7 53.1030.\n\n(4) The amount of financial assurance for decommissioning to be provided must be covered by one or more of the methods described in \u00a7 53.1040 as acceptable to the NRC. A copy of the financial instrument obtained to satisfy the requirements of \u00a7 53.1040 must be submitted to the NRC as part of the application for an OL under this part; however, an applicant for or holder of a COL need not obtain such financial instrument or submit a copy to the Commission except as provided in \u00a7 53.1060(b)."], ["10:10:2.0.1.1.3.7.62.3", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1020 Cost estimates for decommissioning.", "NRC", "", "", "", "Cost estimates for decommissioning must be site-specific. Site-specific decommissioning cost estimates (DCEs) must account for the engineering, labor, equipment, transportation, disposal, and related charges needed to support termination of the license. They must include the costs for decontaminating structures, systems, and components and the site environs; removal of contaminated components and materials from the plant and the site environs; disposal of removed components and materials in appropriate facilities; and any other activities supporting the release of the property and termination of the license. They must also address the approach to annual adjustments required by \u00a7 53.1030. Finally, site-specific DCEs must include plans for adjusting levels of funds assured for decommissioning to demonstrate that a reasonable level of assurance will be provided that funds will be available when needed to cover the cost of decommissioning."], ["10:10:2.0.1.1.3.7.62.4", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1030 Annual adjustments to cost estimates for decommissioning.", "NRC", "", "", "", "Each holder of an OL or COL under this part must annually adjust the cost estimate for decommissioning to account for escalation in labor, energy, and waste burial costs. Licensees may elect to use either a site-specific adjustment factor, approved as part of the plan and associated decommissioning report required by \u00a7 53.1010, in paragraph (a) of this section or the generic adjustment factor in paragraph (b) of this section.\n\n(a) A site-specific adjustment factor must address the estimated contributions and escalation of costs for the following aspects of decommissioning:\n\n(1) Labor, materials, and services;\n\n(2) Energy and waste transportation; and\n\n(3) Radioactive waste burial or other disposition.\n\n(b) A generic adjustment factor must be at least equal to 0.65 L + 0.13 E + 0.22 B, where L and E are escalation factors for labor and energy, respectively, and are to be taken from regional data of U.S. Department of Labor Bureau of Labor Statistics and B is an escalation factor for waste burial and is to be taken from NRC report NUREG-1307, \u201cReport on Waste Burial Charges.\u201d"], ["10:10:2.0.1.1.3.7.62.5", 10, "Energy", "I", "", "53", "PART 53\u2014RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS", "G", "Subpart G\u2014Decommissioning Requirements", "", "\u00a7 53.1040 Methods for providing financial assurance for decommissioning.", "NRC", "", "", "", "Financial assurance for decommissioning is to be provided by the following methods.\n\n(a)  Prepayment.  Prepayment is the deposit made preceding the start of operation or the transfer of a license under \u00a7 53.1570 into an account segregated from licensee assets and outside the administrative control of the licensee and its subsidiaries or affiliates of cash or liquid assets such that the amount of funds would be sufficient to pay decommissioning costs. Prepayment may be in the form of a trust, escrow account, or Government fund with payment by certificate of deposit, deposit of government or other securities, or other method acceptable to the NRC. This trust, escrow account, Government fund, or other type of agreement must be established in writing and maintained at all times in the United States with an entity that is an appropriate State or Federal government agency, or an entity whose operations in which the prepayment deposit is managed are regulated and examined by a Federal or State agency. A licensee that has prepaid funds based on a site-specific cost estimate under \u00a7 53.1020 may take credit for projected earnings on the prepaid decommissioning trust funds, using up to a 2 percent annual real rate of return through the time of termination of the license. A licensee may use a credit of greater than 2 percent if the licensee's rate-setting authority has specifically authorized a higher rate. Actual earnings on existing funds may be used to calculate future fund needs.\n\n(b)  External sinking fund.  An external sinking fund is a fund established and maintained by setting funds aside periodically in an account segregated from licensee assets and outside the administrative control of the licensee and its subsidiaries or affiliates in which the total amount of funds would be sufficient to pay decommissioning costs. An external sinking fund may be in the form of a trust, escrow account, or Government fund, with payment by certificate of deposit, deposit of government or other securities, or other method acceptable to the NRC. This trust, escrow account, Government fund, or other type of agreement must be established in writing and maintained at all times in the United States with an entity that is an appropriate State or Federal government agency, or an entity whose operations in which the external sinking fund is managed are regulated and examined by a Federal or State agency. A licensee that has collected funds based on a site-specific cost estimate under \u00a7 53.1020 may take credit for projected earnings on the external sinking funds using up to a 2 percent annual real rate of return from the time of future funds' collection through the time of termination of the license. A licensee may use a credit of greater than 2 percent if the licensee's rate-setting authority has specifically authorized a higher rate. Actual earnings on existing funds may be used to calculate future fund needs. A licensee whose rates for decommissioning costs cover only a portion of these costs may make use of this method only for the portion of these costs that are collected in one of the manners described in this paragraph (b). This method may be used as the exclusive mechanism relied upon for providing financial assurance for decommissioning in the following circumstances:\n\n(1) By a licensee that recovers, either directly or indirectly, the estimated total cost of decommissioning through rates established by \u201ccost of service\u201d or similar ratemaking regulation. Public utility districts, municipalities, rural electric cooperatives, and State and Federal agencies, including associations of any of the foregoing, that establish their own rates and are able to recover their cost of service allocable to decommissioning, are deemed to satisfy this condition.\n\n(2) By a licensee whose source of revenues for its external sinking fund is a \u201cnon-bypassable charge,\u201d the total amount of which will provide funds estimated to be needed for decommissioning pursuant to \u00a7 53.1020, \u00a7 53.1060, or \u00a7 53.1575.\n\n(c)  A surety method, insurance, or other guarantee method.  (1) These methods guarantee that decommissioning costs will be paid. A surety method may be in the form of a surety bond, or letter of credit. Any surety method or insurance used to provide financial assurance for decommissioning must contain the following conditions:\n\n(i) The surety method or insurance must be open-ended, or, if written for a specified term, such as 5 years, must be renewed automatically, unless 90 days or more prior to the renewal day the issuer notifies the NRC, the beneficiary, and the licensee of its intention not to renew. The surety or insurance must also provide that the full-face amount be paid to the beneficiary automatically prior to the expiration without proof of forfeiture if the licensee fails to provide a replacement acceptable to the NRC within 30 days after receipt of notification of cancellation.\n\n(ii) The surety or insurance must be payable to a trust established for decommissioning costs. The trustee and trust must be acceptable to the NRC. An acceptable trustee includes an appropriate State or Federal government agency or an entity that has the authority to act as a trustee and whose trust operations are regulated and examined by a Federal or State agency.\n\n(2) A parent company guarantee of funds for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix A to 10 CFR part 30.\n\n(3) For commercial companies that issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix C to 10 CFR part 30. For commercial companies that do not issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs may be used if the guarantee and test are as contained in appendix D to 10 CFR part 30. A guarantee by the applicant or licensee may not be used in any situation in which the applicant or licensee has a parent company holding majority control of voting stock of the company.\n\n(d)  Funding method for Federal licensees.  For a Federal licensee, a statement of intent containing a cost estimate for decommissioning and indicating that funds for decommissioning will be obtained when necessary.\n\n(e)  Contractual funding method.  Contractual obligation(s) on the part of a licensee's customer(s), the total amount of which over the duration of the contract(s) will provide the licensee's total share of uncollected funds estimated to be needed for decommissioning pursuant to \u00a7 53.1020, \u00a7 53.1060, or \u00a7 53.1575. To be acceptable to the NRC as a method of decommissioning funding assurance, the terms of the contract(s) must include provisions that the buyer(s) of electricity or other products will pay for the decommissioning obligations specified in the contract(s), notwithstanding the operational status either of the licensed plant to which the contract(s) pertains or force majeure provisions. All proceeds from the contract(s) for decommissioning funding will be deposited to the external sinking fund. The NRC reserves the right to evaluate the terms of any contract(s) and the financial qualifications of the contracting entity or entities offered as assurance for decommissioning funding.\n\n(f)  Other funding mechanisms.  Any other mechanism, or combination of mechanisms, that provides, as determined by the NRC upon its evaluation of the specific circumstances of each licensee submittal, assurance of decommissioning funding equivalent to that provided by the mechanisms specified in paragraphs (a) through (e) of this section. Licensees who do not have sources of funding described in paragraph (b) of this section may use an external sinking fund in combination with a guarantee mechanism, as specified in paragraph (c) of this section, provided that the total amount of funds estimated to be necessary for decommissioning is assured."]], "truncated": false, "filtered_table_rows_count": 227, "expanded_columns": [], "expandable_columns": [], "columns": ["section_id", "title_number", "title_name", "chapter", "subchapter", "part_number", "part_name", "subpart", "subpart_name", "section_number", "section_heading", "agency", "authority", "source_citation", "amendment_citations", "full_text"], "primary_keys": ["section_id"], "units": {}, "query": {"sql": "select section_id, title_number, title_name, chapter, subchapter, part_number, part_name, subpart, subpart_name, section_number, section_heading, agency, authority, source_citation, amendment_citations, full_text from cfr_sections where \"part_number\" = :p0 and \"title_number\" = :p1 order by section_id limit 101", "params": {"p0": "53", "p1": "10"}}, "facet_results": {"title_number": {"name": "title_number", "type": "column", "hideable": false, "toggle_url": "/openregs/cfr_sections.json?part_number=53&title_number=10", "results": [{"value": 10, "label": 10, "count": 227, "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53", "selected": true}], "truncated": false}, "agency": {"name": "agency", "type": "column", "hideable": false, "toggle_url": "/openregs/cfr_sections.json?part_number=53&title_number=10", "results": [{"value": "NRC", "label": "NRC", "count": 227, "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53&title_number=10&agency=NRC", "selected": false}], "truncated": false}, "part_number": {"name": "part_number", "type": "column", "hideable": false, "toggle_url": "/openregs/cfr_sections.json?part_number=53&title_number=10", "results": [{"value": "53", "label": "53", "count": 227, "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?title_number=10", "selected": true}], "truncated": false}}, "suggested_facets": [{"name": "subpart", "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53&title_number=10&_facet=subpart"}, {"name": "subpart_name", "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53&title_number=10&_facet=subpart_name"}, {"name": "amendment_citations", "toggle_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53&title_number=10&_facet=amendment_citations"}], "next": "10~3A10~3A2~2E0~2E1~2E1~2E3~2E7~2E62~2E5,10~3A10~3A2~2E0~2E1~2E1~2E3~2E7~2E62~2E5", "next_url": "https://www.pawtectors.org/openregs/cfr_sections.json?part_number=53&title_number=10&_next=10~3A10~3A2~2E0~2E1~2E1~2E3~2E7~2E62~2E5%2C10~3A10~3A2~2E0~2E1~2E1~2E3~2E7~2E62~2E5&_sort=section_id", "private": false, "allow_execute_sql": true, "query_ms": 144.2975359968841, "source": "Federal Register API & Regulations.gov API", "source_url": "https://www.federalregister.gov/developers/api/v1", "license": "Public Domain (U.S. Government data)", "license_url": "https://www.regulations.gov/faq"}