美国核管会,10 CFR Part 50, Appendix A,核电厂设计总则
核电站仪控设计管理标准和规范学习
IAEA安全标准
三个层次安全基础（Safety Fundamentals）安全要求（Safety Requirements）安全导则（Safety Guides）五个领域基本安全 General Safety (GS) - All committees核安全 Nuclear Safety (NS) - NUSSC辐射安全 Radiation Safety (RS) - RASSC运输安全 Transport Safety (TS) - TRANSSC放废安全 Waste Safety (WS) - WASSC
国家核安全法规和导则
HAF 602-2007 民用核安全设备无损检验人员资格管理规定HAF 603-2007 民用核安全设备焊工焊接操作工资格管理规定HAF 604-2007 进口民用核安全设备监督管理规定导则（HAD）HAD 002/01-1989 核动力营运单位的应急准备HAD 003/01-10 核电厂质量保证系列导则HAD 101/01-12 核电厂厂址选择系列导则HAD 102/01-17 核电厂设计系列导则HAD 103/01-11 核电厂运行系列导则HAD 301/02-1998 乏燃料贮存设施的设计HAD 401/01-06 核电厂放射性废物系列导则HAD 501/02-1998 核动力厂实物保护导则技术文件（HAF J）HAF.J0053-1995 核设备抗震鉴定试验指南
IAEA安全标准
安全导则（Safety Guides）IAEA-NS-G-1.1-1.12 核动力厂设计IAEA-NS-G-2.1-2.15 核动力厂运行IAEA-NS-G-3.1-3.6 核装置的厂址评价IAEA-GS- G- 3.1 设施和活动的管理体系应用(QA)安全报告系列（Safety Reports）IAEA技术报告N.384：Verification and Validation of Software Related to Nuclear Power Plant Instrumentation and ControlIAEA技术报告N.387： Modern Instrumentation and Control for Nuclear Power Planter Plants
美国核管会CRP17.5中文翻译稿
美国核管理委员会标准审查大纲（NUREG-0800）17.5 质量保证大纲概述-设计证明书、早期厂址许可和新许可证申请者审查责任主审—负责质量保证（QA）的机构副审—无Ⅰ.审查范围质量保证人员审评由申请者提交的关于设计证明书（DC）、建造和运行联合许可证（COL）、早期厂址许可（ESP）、建造许可（CP）及运行许可证（OL）的质量保证大纲概述（QAPDs）。
按照本标准审查大纲（SRP）适用的章节对申请者提交的关于设计证明书、建造和运行联合许可证、早期厂址许可、建造许可及运行许可证的质量保证大纲概述进行审评。
由设计证明书申请者提交的质量保证大纲概述（QAPD）可以是质量保证专题报告或部分安全分析报告（SAR）。
由设计证明书申请者提交的质量保证大纲概述只会论及支持设计证明书的设计质量保证活动。
该质量保证大纲概述不会论及建造开始后发生的建造和设计质量保证活动。
在NRC批准DC前，NRC审评由设计证明书申请者提交的质量保证大纲概述。
由建造和运行联合许可证申请者提交的质量保证大纲概述适用于设施寿期的所有阶段，包括设计、建造以及运行。
建造和运行阶段的质量保证活动可在单独的质量保证大纲概述中论述。
在运行阶段，建造和运行联合许可证申请者可以参考经美国核管会核准的质量保证大纲概述。
但是，将依据在提交申请前6个月生效的SRP审查申请书。
早期厂址许可申请者提交的质量保证大纲概述适用于厂址适宜性质量保证活动，并由美国核管会在核发早期厂址许可前审评。
建造许可申请者提交的质量保证大纲概述适用于所有设计和建造质量保证活动，并由美国核管会在核发建造许可前审评。
运行许可证申请者提交的质量保证大纲概述适用于运行阶段质量活动，并由美国核管会在核发运行许可证前审评。
基于美国国家标准学会（ANSI）N45.2“核电厂质量保证大纲要求”及其子标准，标准审查大纲17.1节和17.2节规定质量保证大纲的审查导则。
标准审查大纲17.3节规定基于美国机械工程师协会（ASME）NQA-1“核设施质量保证大纲”和NQA-2“核设施申请的质量保证要求”编写的质量保证大纲概述的审查导则。
10CFR21—核能电厂及核燃料再处理厂品质保证准则美国联邦法规第10篇第21章--英文版
PART 21 REPORTING OF DEFECTS AND NONCOMPLIANCEGENERAL PROVISIONS (1)§21.1P URPOSE. (1)§21.2S COPE. (1)§21.3D EFINITIONS. (2)§21.4I NTERPRETATIONS. (5)§21.5C OMMUNICATIONS. (5)§21.6P OSTING R EQUIREMENTS. (5)§21.7E XEMPTIONS. (6)§21.8I NFORMATION COLLECTION REQUIREMENTS:OMB APPROVAL. (6)NOTIFICATION (6)§21.21N OTIFICATION OF FAILURE TO COMPLY OR EXISTENCE OF A DEFECT AND ITS EVALUATION. (6)PROCUREMENT DOCUMENTS (8)§21.31P ROCUREMENT DOCUMENTS. (8)INSPECTIONS, RECORDS (9)§21.41I NSPECTIONS. (9)§21.51M AINTENANCE AND INSPECTION OF RECORDS. (9)ENFORCEMENT (9)§21.61F AILURE TO NOTIFY. (9)§21.62C RIMINAL PENALTIES. (10)G ENERAL P ROVISIONS§ 21.1 Purpose.The regulations in this part establish procedure and requirements for implementation of section 206 of the Energy Reorganization Act of 1974. That section requires any individual director or responsible officer of a firm constructing, owning, operating or supplying the components of any facility or activity which is licensed or otherwise regulated pursuant to the Atomic Energy Act of 1954, as amended, or the Energy Reorganization Act of 1974, who obtains information reasonably indicating: (a) That the facility, activity or basic component supplied to such facility or activity fails to comply with the Atomic energy Act of 1954, as amended, or any applicable rule, regulation, order, or license of the Commission relating to substantial safety hazards or (b) that the facility, activity, or basic component supplied to such facility or activity contains defects, which could create a substantial safety hazard, to immediately notify the Commission of such failure to comply or such defect, unless he has actual knowledge that the Commission has been adequately informed of such defect or failure to comply.§ 21.2 Scope.(a) The regulations in this part apply, except as specifically provided otherwise in parts 31, 34, 35, 39, 40, 60, 61, 70, or part 72 of this chapter, to each individual, partnership, corporation, or other entity licensed pursuant to the regulations in this chapter to possess, use, or transfer within the United States source material, byproduct material, special nuclear material, and/or spent fuel and high level radioactive waste, or to construct, manufacture, possess, own, operate or transfer within the United States, any production or utilization facility or independent spent fuel storage installation (ISFSI) or monitored retrievable storage installation (MRS); and to eachdirector and responsible officer of such a licensee. The regulations in this part apply also to each individual, corporation, partnership or other entity doing business within the United States, and each director and responsible officer of such organization, that constructs a production or utilization facility licensed for manufacture, construction, or operation pursuant to part 50 of this chapter, an ISFSI for the storage of spent fuel licensed pursuant to part 72 of this chapter, a MRS for the storage of spent fuel or high level radioactive waste pursuant to part 72 of this chapter, or a geologic repository for the disposal of high-level radioactive waste under part 60 of this chapter; or supplies basic components for a facility or activity licensed, other than for export, under parts 30, 40, 50, 60, 61, 70, 71, or part 72 of this chapter.(b) For persons licensed to construct a facility under a construction permit issued under §50.23 of this chapter, evaluation of potential defects and failures to comply and reporting of defects and failures to comply under § 50.55(e) of this chapter satisfies each person’s evaluation, notification, and reporting obligation to report defects and failures to comply under this part and the responsibility of individual directors and responsible officers of such licensees to report defects under section 206 of the Energy Reorganization Act of 1974.(c) For persons licensed to operate a nuclear power plant under part 50 of this chapter, evaluation of potential defects and appropriate reporting of defects under §§ 50.72, 50.73 or § 73.71 of this chapter satisfies each person’s evaluation, notification, and reporting obligation to report defects under this part and the responsibility of individual directors and responsible officers of such licensees to report defects under section 206 of the Energy Reorganization Act of 1974.(d) Nothing in these regulations should be deemed to preclude either an individual, a manufacturer, or a supplier of a commercial grade item (as defined in § 21.3) not subject to the regulations in this part from reporting to the Commission, a known or suspected defect or failure to comply and, as authorized by law, the identity of anyone so reporting will be withheld from disclosure. NRC regional offices and headquarters will accept collect telephone calls from individuals who wish to speak to NRC representatives concerning nuclear safety-related problems. The location and telephone numbers of the four regions (answered during regular working hours) are listed in appendix D to part 20 of this chapter. The telephone number of the NRC Operations Center (answered 24 hours a day—including holidays) is (301) 951-0550.§ 21.3 Definitions.As used in this part:Basic component. (1)(i) When applied to nuclear power plants licensed pursuant to 10 CFR 50 of this chapter, basic component means a structure, system, component or part thereof that affects its safety function necessary to assure:(A) The integrity of the reactor coolant pressure boundary;(B) The capability to shut down the reactor and maintain it in a safe shutdown condition; or(C) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1) or § 100.11 of this chapter, as applicable.(ii) Basic components are items designed and manufactured under a quality assurance program complying with 10 CFR Part 50, Appendix B, or commercial grade items which have successfully completed the dedication process.(2) When applied to other facilities and when applied to other activities licensed pursuant to Parts 30, 40, 50 (other than nuclear power plants), 60, 61, 70, 71, or 72 of this chapter, basic component means a structure, system, or component, or part thereof that affects their safety function, that is directly procured by the licensee of a facility or activity subject to the regulations in this part and in which a defect or failure to comply with any applicable regulation in this chapter, order, or license issued by the Commission could create a substantial safety hazard.(3) In all cases, basic component includes safety related design, analysis, inspection, testing, fabrication, replacement parts, or consulting services that are associated with the component hardware whether these services are performed by the component supplier or others.Commercial grade item. (1) When applied to nuclear power plants licensed pursuant to 10 CFR Part 50, commercial grade item means a structure, system, or component, or part thereof that affects its safety function, that was not designed and manufactured as a basic component. Commercial grade items do not include items where the design and manufacturing process require in-process inspections and verifications to ensure that defects or failures to comply are identified and corrected (i.e., one or more critical characteristics of the item cannot be verified).(2) When applied to facilities and activities licensed pursuant to 10 CFR Parts 30, 40, 50 (other than nuclear power plants), 60, 61, 70, 71, or 72, commercial grade item means an item that is:(i) Not subject to design or specification requirements that are unique to those facilities or activities;(ii) Used in applications other than those facilities or activities; and(iii) To be ordered from the manufacturer/ supplier on the basis of specifications set forth in the manufacturer’s published product description (for example, a catalog).Commission means the Nuclear Regulatory Commission or its duly authorized representatives.Constructing or construction means the analysis, design, manufacture, fabrication, placement, erection, installation, modification, inspection, or testing of a facility or activity which is subject to the regulations in this part and consulting services related to the facility or activity that are safety related.Critical characteristics. When applied to nuclear power plants licensed pursuant to 10 CFR Part 50, critical characteristics are those important design, material, and performance characteristics of a commercial grade item that, once verified, will provide reasonable assurance that the item will perform its intended safety function.Dedication. (1) When applied to nuclear power plants licensed pursuant to 10 CFR Part 50, dedication is an acceptance process undertaken to provide reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety function and, in this respect, is deemed equivalent to an item designed and manufactured under a 10 CFR Part 50, Appendix B, quality assurance program. This assurance is achieved by identifying the critical characteristics of the item and verifying their acceptability by inspections, tests, or analyses performed by the purchaser or third-party dedicating entity after delivery, supplemented as necessary by one or more of the following commercial grade surveys; product inspections or witness at holdpoints at the manufacturer’s facility, and analysis of historical records for acceptable performance. In all cases, the dedication process must be conducted in accordance with the applicable provisions of 10 CFR Part 50, Appendix B. The process is considered complete when the item is designated for use as a basic component.(2) When applied to facilities and activities licensed pursuant to 10 CFR Parts 30, 40, 50 (other than nuclear power plants), 60, 61, 70, 71, or 72, dedication occurs after receipt when that item is designated for use as a basic component.Dedicating entity. When applied to nuclear power plants licensed pursuant to 10 CFR Part 50, dedicating entity means the organization that performs the dedication process. Dedication may be performed by the manufacturer of the item, a third-party dedicating entity, or the licensee itself. The dedicating entity, pursuant to § 21.21(c) of this part, is responsible for identifying and evaluating deviations, reporting defects and failures to comply for the dedicated item, and maintaining auditable records of the dedication process.Defect means:(1) A deviation in a basic component delivered to a purchaser for use in a facility or an activity subject to the regulations in this part if, on the basis of an evaluation, the deviation could create a substantial safety hazard; or(2) The installation, use, or operation of a basic component containing a defect as defined in this section; or(3) A deviation in a portion of a facility subject to the construction permit or manufacturing licensing requirements of Part 50 of this chapter provided the deviation could, on the basis of an evaluation, create a substantial safety hazard and the portion of the facility containing the deviation has been offered to the purchaser for acceptance; or(4) A condition or circumstance involving a basic component that could contribute to the exceeding of a safety limit, as defined in the technical specifications of a license for operation issued pursuant to Part 50 of this chapter.Deviation means a departure from the technical requirements included in a procurement document.Director means an individual, appointed or elected according to law, who is authorized to manage and direct the affairs of a corporation, partnership or other entity. In the case of an individual proprietorship, “director’ means the individual.Discovery means the completion of the documentation first identifying the existence of a deviation or failure to comply potentially associated with a substantial safety hazard within the evaluation procedures discussed in § 21.21(a).Evaluation means the process of determining whether a particular deviation could create a substantial hazard or determining whether a failure to comply is associated with a substantial safety hazard.Notification means the telephonic communication to the NRC Operations Center or written transmittal of information to the NRC Document Control Desk.Operating or operation means the operation of a facility or the conduct of a licensed activity which is subject to the regulations in this part and consulting services related to operations that are safety related.Procurement document means a contract that defines the requirements which facilities or basic components must meet in order to be considered acceptable by the purchaser.Responsible officer means the president, vice-president or other individual in the organization of a corporation, partnership, or other entity who is vested with executive authority over activities subject to this part.Substantial safety hazard means a loss of safety function to the extent that there is major reduction in the degree of protection provided to public health and safety for any facility or activity licensed, other than for export, pursuant to Parts 30, 40, 50, 60, 61, 70, 71, or 72 of this chapter.Supplying or supplies means contractually responsible for a basic component used or to be used in a facility or activity which is subject to the regulations of this part.§ 21.4 Interpretations.Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other that a written interpretation by the General Counsel will be recognized to be binding upon the Commission.§ 21.5 Communications.Except where otherwise specified in this part, all written communications and reports concerning the regulations in this part must be addressed to the Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555. In the case of a licensee, a copy must also be sent to the appropriate Regional Administrator at the address specified in appendix D to part 20 of this chapter.§ 21.6 Posting Requirements.(a)(1) Each individual, partnership, corporation or other entity subject to the regulations in this part, shall post current copies of—(i) The regulations in this part.(ii) Section 206 of the Energy Reorganization Act of 1974; and(iii) Procedures adopted pursuant to the regulations in this part.(2) These documents must be posted in a conspicuous position on any premises within the United States where the activities subject to this part are conducted.(b) If posting of the regulations in this part or the procedures adopted pursuant to the regulations in this part is not practicable, the licensee or firm subject to the regulations in this part may, in addition to posting section 206, post a notice which describes the regulations/procedures, including the name of the individual to whom reports may be made, and states where they may be examined.(c) The effective date of this section has been deferred until January 6, 1978.§ 21.7 Exemptions.The Commission may, upon application of any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Suppliers of commercial grade items are exempt from the provisions of this part to the extent that they supply commercial grade items.§ 21.8 Information collection requirements: OMB approval.(a) The Nuclear Regulatory Commission 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 of 1980 (44 U.S.C. 3501 et seq.). OMB has approved the information collection requirements contained in this part under control number 3150-0035.(b) The approved information collection requirements contained in this part appear in §§21.21 and 21.51.N OTIFICATION§ 21.21 Notification of failure to comply or existence of a defect and its evaluation.(a) Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part must adopt appropriate procedures to—(1) Evaluate deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazards as soon as practicable, and, except as provided in paragraph (a)(2) of this section, in all case within 60 days of discovery, in order to identify a reportable defect or failure to comply that could create a substantial safety hazard, were it to remain uncorrected, and(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 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 § 21.21(c)(5). The interim report should describe the deviation or failure to comply that is being evaluated and should also state when the evaluation will be completed. This interim report must be submitted in writing within 60 days of discovery of the deviation or failure to comply.(3) Ensure that a director or responsible officer subject to the regulations of this part is informed as soon as practicable, and, in all cases, within the 5 working days after completion of the evaluation described in § 21.21(a)(1) or § 21.21(a)(2) if the construction or operation of a facility or activity, or a basic component supplied for such facility or activity—(i) Fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable rule, regulation, order, or license of the Commission relating to a substantial safety hazard, or (ii) Contains a defect.(b) If the deviation or failure to comply is discovered by a supplier of basic components, or services associated with basic components, and the supplier determines that it does not have the capability to perform the evaluation to determine if a defect exists, then the supplier must inform the purchasers of affected licensees within five working days of this determination so that the purchasers or affected licensees may evaluate the deviation or failure to comply, pursuant to §21.21(a).(c) A dedicating entity is responsible for—(1) Identifying and evaluating deviations and reporting defects and failures to comply associated with substantial safety hazards for dedicated items; and(2) Maintaining auditable records for the dedication process.(d)(1) A director or responsible officer subject to the regulations of this part or a person designated under § 21.21(c)(5) must notify the Commission when he or she obtains information reasonably indicating a failure to comply or a defect affecting—(i) The construction or operation of a facility or an activity within the United States that is subject to the licensing requirements under parts 30, 40, 50, 60, 61, 70, 71, or 72 of this chapter and that is within his or her organization’s responsibility; or(ii) A basic component that is within his or her organization’s responsibility and is supplied for a facility or an activity within the United States that is subject to the licensing requirements under parts 30, 40, 50, 60, 61, 70, 71, or 72 of this chapter.(2) The notification to NRC of a failure to comply or of a defect under paragraph (c)(1) of this section and the evaluation of a failure to comply or a defect under paragraphs (a)(1) and (a)(2) of this section, are not required if the director or responsible officer has actual knowledge that the Commission has been notified in writing of the defect or the failure to comply.(3) Notification required by paragraph (c)(1) of this section must be made as follows—(i) Initial notification by facsimile, which is the preferred method of notification, to the NRC Operations Center at 301-492-8187 or by telephone at 301-951-0550 within two days following receipt of information by the director or responsible corporate officer under paragraph (a)(1) of this section, on the identification of a defect or a failure to comply. Verification that the facsimile has been received should be made by calling the NRC Operations Center. This paragraph does not apply to interim reports described in § 21.21(a)(2).(ii) Written notification to the NRC at the address specified in § 21.5 within 30 days following receipt of information by the director or responsible corporate officer under paragraph(a)(3) of this section, on the identification of a defect or a failure to comply.(4) The written report required by this paragraph shall include, but need not be limited to, the following information, to the extent known:(i) Name and address of the individual or individuals informing the Commission.(ii) Identification of the facility, the activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect.(iii) Identification of the firm constructing the facility or supplying the basic component which fails to comply or contains a defect.(iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.(v) The date on which the information of such defect or failure to comply was obtained.(vi) In the case of a basic component which contains a defect or fails to comply, the number and location of all such components in use at, supplied for, or being supplied for one or more facilities or activities subject to the regulations in this part.(vii) 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, is being, or will be taken to complete the action.(viii) 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 purchasers or licensees.(5) The director or responsible officer may authorize an individual to provide the notification required by this paragraph, provided that, this shall not relieve the director or responsible officer of his or her responsibility under this paragraph.(e) Individuals subject to this party may be required by the Commission to supply additional information related to a defect or failure to comply. Commission action to obtain additional information may be based on reports of defects from other reporting entities.P ROCUREMENT D OCUMENTS§ 21.31 Procurement documents.Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall assure that each procurement document for facility, or a basic component issued by him, her or it on or after January 6, 1978 specifies, when applicable, that the provisions of 10 CFR Part 21 apply.I NSPECTIONS,R ECORDS§ 21.41 Inspections.Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall permit the Commission to inspect its records, premises, activities, and basic components as necessary to accomplish the purposes of this part.§ 21.51 Maintenance and inspection of records.(a) Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall prepare and maintain records necessary to accomplish the purposes of this part, specifically—(1) Retain evaluations of all deviations and failures to comply for a minimum of five years after the date of the evaluation;(2) Suppliers of basic components must retain any notifications sent to purchasers and affected licensees for a minimum of five years after the date of the notification;(3) Suppliers of basic components must retrain a record of the purchasers of basic components for 10 years after delivery of the basic component or service associated with a basic component.(b) Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall permit the Commission the opportunity to inspect records pertaining to basic components that relate to the identification and evaluation of deviations and failures to comply, including any advice given to purchasers or licensees on the placement, erection, installation, operation, maintenance, modification, or inspection of a basic component.E NFORCEMENT§ 21.61 Failure to notify.(a) Any director or responsible officer of an entity (including dedicating entity) that is not otherwise subject to the deliberate misconduct provisions of this chapter but is subject to the regulations in this part who knowingly and consciously fails to provide the notice required by § 21.21 shall be subject to a civil penalty equal to the amount provided by section 234 of the Atomic Energy Act of 1954, as amended.(b) Any NRC licensee subject to the regulations in this part who fails to provide the notice required by § 21.21 or otherwise fails to comply with the applicable requirements of this part shall be subject to a civil penalty as provided by section 234 of the Atomic Energy Act of 1954, as amended.(c) The dedicating entity, pursuant to § 21.21(c) of this part, is responsible for identifying and evaluating deviations, reporting defects and failures to comply for the dedicated item, and maintaining auditable records of the dedication process. NRC enforcement action can be taken for failure to identify and evaluate deviations, failure to report defects and failures to comply, or failure to maintain auditable records.§ 21.62 Criminal penalties.(a) Section 223 of the Atomic Energy Act of 1954, as amended, 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 21 issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.(b) The regulations in part 21 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 21.1, 21.2, 21.3, 21.4, 21.5, 21.7, 21.8, 21.61, and 21.62.[57 FR 55071, Nov. 24, 1992]。
美国核安全法规介绍
美国核安全法规介绍一、美国核电法规体系的五个层次：二、美国核电法规和标准简介2.1 原子能法（第一层次）原子能法，美国国会参众两院于1954年批准并公布，共有303条，分成20章。
原子能法是美国对原子能的和平利用和军事用途管理的根本依据。
2.2 联邦法规（第二层次）联邦法规，美国联邦法规由美国核管理委员会（NRC）发布；第10部分是“能源”，它规定了和平利用原子能通用的和特殊的原则和准则，它在美国具有法律效力。
第10部分“能源”与核电厂设计有关的部分主要有：10CFR50“生产和应用设施的执照发放”的附录（15个）2.3 美国核管理委员会的管理导则（第三层次）美国核管理委员会的管理导则，美国核管理委员会制定了一整套的管理导则（RG）它提供了符合法规要求的指导和可行的解决办法。
按照不同内容，将这些导则分为10个部分，涉及核电厂的内容编为第一部分，即RG.1。
如：RG.1.28《质量保证大纲要求（设计和建造）》；RG.1.38《轻水堆核电厂各物项的包装、运输、接受、贮存和装卸的质量保证要求》；RG.1.64《核电厂设计的质量保证要求》；RG.1.70《核电厂安全分析报告的标准格式和内容》等。
管理导则的其它部分为研究和试验反应堆、核燃料和物料设备、环境和厂址以及职业保健等。
2.4 美国核管理委员会的技术文件（NUREG）（第四层次）▲NUREG文件：美国核管理委员会下设的反应堆管理局负责编制的技术文件；▲NUREG/CR文件：委托各种研究机构完成的技术文件。
NUREG文件和NUREG/CR文件属于建议性的参考文件；有时NUREG文件与R.G具有同样的作用：如“NUREG-0800”是《核电厂安全分析报告的标准审查大纲》，这是NRC 对申请者按照“R.G.1.70”《核电厂安全分析报告的标准格式和内容》要求编写的“初步/最终安全分析报告”进行审查的指导性文件。
我国的国家核安全局也是参照该技术文件审查核电站的安全分析报告。
核安全法规和导则介绍
2.2 我国的核安全质保法规HAF003和导则
－我国于1984年成立国家核安全局，1986年该局参
考IAEA50-C-QA《核电厂安全质量保证实施法规》 （1978）法规发布了HAF0400《核电厂质量保证安 全规定》，其后陆续发布了10个安全导则。1991 年根据IAEA 50-C/SG-QA(1988)，发布HAF0400 （91）。2001年国家核安全局对核安全法规和导 则进行了补充和修订，重新编号HAF/HAD 003系列 法规、导则(见表2）。
－核电站上述安全措施是由大量与核安全相关的物项、系统 和构筑物构成的，尽管这些物项、系统和构筑物在设计上 遵循了核安全设计法规、技术规范和标准要求。但如果在 设计、制造中不遵循核安全质保法规要求，不建立质量保 证大纲和实施所规定的各项质量保证活动，则物项、系统 和构筑物质量无法保证，安全目标也难于实现。 －国际原子能机构的核安全法规除IAEA 50-C-G 政府机构， 50-C-S 选址，50-C-D 设计和50-C-O 运行4个法规外，还 颁布了IAEA 50-C-Q安全质量保证法规。这5个安全法规共 同构成核电站安全要求的基础。由此可见，安全质量保证 法规是核安全法规的一个重要的不可分割的组成部分。核 设备的质量保证是核电站一项重要的安全措施。 •
－我国核安全质保法规HAF003系列是从IAEA
50-C-QA
（1988）法规等效采用的。而 IAEA 50-C-QA（1996）法 规包括了IAEA 50-C-QA（1988）法规的全部要求并有所扩 展。
－国内取得国家核安全局核设备设计、制造许可证的制造厂 已按HAF003系列核安全法规和导则标准制定和实施了质量 保证大纲。这就为在C3/C4项目设备设计制造中实施IAEA 50-C-QA（1996）法规提供了较好的基础。
10CFR50附录B英文版
Appendix B to 10CFR Part 50Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Introduction. Every applicant for a construction permit is required by the provisions of § 50.34 to include in itspreliminary safety analysis report a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. Every applicant for an operating license is required to include, in its final safety analysis report, information pertaining to the managerial and administrative controls to be used to assure safe operation. Every applicant for a combined license under part 52 of this chapter is required by the provisions of § 52.79 of this chapter to include in its final safety analysis report a description of the quality assurance applied to the design, and to be applied to the fabrication, construction, and testing of the structures, systems, and components of the facility and to the managerial and administrative controls to be used to assure safe operation. For applications submitted after September 27, 2007, every applicant for an early site permit under part 52 of this chapter is required by the provisions of § 52.17 of this chapter to include in its site safety analysis report a description of the quality assurance program applied to site activities related to the design,fabrication, construction, and testing of the structures, systems, and components of a facility or facilities that may be constructed on the site. Every applicant for a design approval or design certification under part 52 of this chapter is required by the provisions of 10 CFR 52.137 and 52.47, respectively, to include in its final safety analysis report adescription of the quality assurance program applied to the design of the structures, systems, and components of the facility. Every applicant for a manufacturing license under part 52 of this chapter is required by the provisions of 10 CFR 52.157 to include in its final safety analysis report a description of the quality assurance program applied to the design, and to be applied to the manufacture of, the structures, systems, and components of the reactor. Nuclear power plants and fuel reprocessing plants include structures, systems, and components that prevent or mitigate the consequences of postulated accidents that structures, systems, and components. The pertinent requirements of this appendix apply to all activities affecting the safetyrelated functions of those structures, systems, and components; these activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.As used in this appendix, "quality assurance" comprises all those planned and systematic actions necessary to provideadequate confidence that a structure, system, or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance 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.I. OrganizationThe applicant (1) shall be responsible for the establishment and execution of the quality assurance program. The applicant may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part thereof, but shall retain responsibility for the quality assurance program. The authority and duties of persons and organizations performing activities affecting the safety-related functions of structures, systems, and components shall be clearly established and delineated in writing. These activities include both the performing functions of attaining quality objectives and the quality assurance functions. The quality assurancefunctions are those of (1) assuring that an appropriate quality assurance program is established and effectively executed; and (2) verifying, such as by checking, auditing, and inspecting, that activities affecting the safetyrelatedfunctions have been correctly performed. The persons and organizations performing quality assurance functions shallhave sufficient authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions. There persons and organizations performing quality assurance functions shall report to a management level so that the required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are provided. Because of thecould cause undue risk to the health and safety of the public. This appendix establishes quality assurance requirements for the design, manufacture, construction, and operation of thosemany variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms, provided that the persons and organizations assigned the quality assurance functions have the required authority and organizational freedom. Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portionof the quality assurance program at any location where activities subject to this appendix are being performed, shall have direct access to the levels of management necessary to perform this function.II. Quality Assurance ProgramThe applicant shall establish at the earliest practicable time, consistent with the schedule for accomplishing the activities, a quality assurance program which complies with the requirements of this appendix. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions. The applicant shall identify the structures, systems, and components to be covered by the quality assurance program and the major organizations participating in the program, together with the designated functions of these organizations. The quality assurance program shall provide control over activities affecting the quality of the identified structures, systems, and components, to an extent consistent with their importance to safety. Activities affecting quality shall be accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanness; and assurance that all prerequisites for the given activity have been satisfied. The program shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test. The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained. The applicant shall regularly review the status and adequacy of the quality assurance program. Management of other organizations participating in the quality assurance program shall regularly review the status and adequacy of that part of the quality assurance program which they are executing.III. Design ControlMeasures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from suchstandards are controlled.Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.Measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces.The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. The verifying or checking process shall be performed by individuals or groups other than those who performed the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, it shall include suitable qualifications testing of a prototype unit under the most adverse design conditions. Design control measures shall be applied to items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials; accessibility for inservice inspection, maintenance, and repair; and delineation of acceptance criteria for inspections and tests.Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another responsible organization.IV. Procurement Document ControlMeasures shall be established to assure that applicable regulatory requirements, design bases, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services, whether purchased by the applicant or by its contractors or subcontractors. To the extent necessary, procurement documents shall require contractors or subcontractors to provide a quality assurance program consistent with the pertinent provisions of this appendix.V. Instructions, Procedures, and DrawingsActivities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate tohe circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.VI. Document ControlMeasures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel and are distributed to and used at the location where the prescribed activity is performed. Changes to documents shall be reviewed and approved by the same organizations that performed the original review and approval unless the applicant designates another responsible organization.VII. Control of Purchased Material, Equipment, and ServicesMeasures shall be established to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures shall include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products upon delivery. Documentary evidence that material and equipment conform to the procurement requirements shall be available at the nuclear powerplant or fuel reprocessing plant site prior to installation or use of such material and equipment. This documentary evidence shall be retained at the nuclear powerplant or fuel reprocessing plant site and shall be sufficient to identify the specific requirements, such as codes, standards, or specifications, met by the purchased material and equipment. The effectiveness of the control of quality by contractors and subcontractors shall be assessed by the applicant or designee at intervals consistent with the importance, complexity, and quantity of the product or services.VIII. Identification and Control of Materials, Parts, and ComponentsMeasures shall be established for the identification and control of materials, parts, and components, including partially fabricated assemblies. These measures shall assure that identification of the item is maintained by heat number, part number, serial number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation, and use of the item. These identification and control measures shall be designed to prevent tthe use of incorrect or defective material, parts, and components.IX. Control of Special ProcessesMeasures shall be established to assure that special processes, including welding, heat treating, and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.X. InspectionA program for inspection of activities affecting quality shall be established and executed by or for the organization performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. Such inspection shall be performed by individuals other than those who performed the activity being inspected. Examinations, measurements, or tests of material or products processed shall be performed for each work operation where necessary to assure quality. If inspection of processed material or products is impossible or disadvantageous, indirect control by monitoring processing methods, equipment, and personnel shall be provided. Both inspection and process monitoring shall be provided when control is inadequate without both. If mandatory inspection hold points, which require witnessing or inspecting by the applicant's designated representative and beyond which work shall not proceed without the consent of its designated representative are required, the specific hold points shall be indicated in appropriate documents.XI. Test ControlA test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate, proof tests prior to installation, preoperational tests, and operational tests during nuclear power plant or fuel reprocessing plant operation, of structures, systems, and components. Test procedures shall include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. Test results shall be documented and evaluated to assure that test requirements have been satisfied.XII. Control of Measuring and Test EquipmentMeasures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits. XIII. Handling, Storage and ShippingMeasures shall be established to control the handling, storage, shipping, cleaning and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, specific moisture content levels, and temperature levels, shall be specified and provided.XIV. Inspection, Test, and Operating StatusMeasures shall be established to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed upon individual items of the nuclear power plant or fuel reprocessing plant. These measures shall provide for the identification of items which have satisfactorily passed required inspections and tests, where necessaryto preclude inadvertent bypassing of such inspections and tests. Measures shall also be established for indicating the operating status of structures, systems, and components of the nuclear power plant or fuel reprocessing plant, such as by tagging valves and switches, to prevent inadvertent operation.XV. Nonconforming Materials, Parts, or ComponentsMeasures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures.XVI. Corrective ActionMeasures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.XVII. Quality Assurance RecordsSufficient records shall be maintained to furnish evidence of activities affecting quality. The records shall include at least the following: Operating logs and the results of reviews, inspections, tests, audits, monitoring of work performance, and materials analyses. The records shall also include closely-related data such as qualifications of personnel, procedures, and equipment. Inspection and test records shall, as a minimum, identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted. Records shall be identifiable and retrievable. Consistent with applicableregulatory requirements, the applicant shall establish requirements concerning record retention, such as duration, location, and assigned responsibility.XVIII. AuditsA comprehensive system of planned and periodic audits shall be carried out to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits shall be performed in accordance with the written procedures or check lists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audit results shall be documented and reviewed by management having responsibility in the area audited. Followup action, including reaudit of deficient areas, shall be taken where indicated.[35 FR 10499, June 27, 1970, as amended at 36 FR 18301, Sept. 11, 1971; 40 FR 3210D, Jan. 20, 1975; 72 FR 49505, Aug. 28, 2007] 1 While the term "applicant" is used in these criteria, the requirements are, of course, applicable after such a person has received a license to construct and operate a nuclear power plant or a fuel reprocessing plant or has received an early site permit, design approval, design certification, or manufacturing license, as applicable. These criteria will also be used for guidance in evaluating the adequacy of quality assurance programs in use by holders of construction permits, operating licenses, early site permits, design approvals, combined licenses, and manufacturing licenses.。
美国核安全法规
美国核安全法规标准介绍一、美国核电法规体系的五个层次：二、美国核电法规和标准简介2.1 原子能法（第一层次）（放射性污染防治法）原子能法，美国国会参众两院于1954年批准并公布，共有303条，分成20章。
原子能法是美国对原子能的和平利用和军事用途管理的根本依据。
2.2 联邦法规（第二层次）（HAF）联邦法规，美国联邦法规由美国核管理委员会（NRC）发布；第10部分是“能源”，它规定了和平利用原子能通用的和特殊的原则和准则，它在美国具有法律效力。
第10部分“能源”与核电厂设计有关的部分主要有：10CFR50“生产和应用设施的执照发放”的附录（15个）2.3 美国核管理委员会的管理导则（第三层次）（HAD）美国核管理委员会的管理导则，美国核管理委员会制定了一整套的管理导则（RG）它提供了符合法规要求的指导和可行的解决办法。
按照不同内容，将这些导则分为10个部分，涉及核电厂的内容编为第一部分，即RG.1。
如：RG.1.28《质量保证大纲要求（设计和建造）》；RG.1.38《轻水堆核电厂各物项的包装、运输、接受、贮存和装卸的质量保证要求》；RG.1.64《核电厂设计的质量保证要求》；RG.1.70《核电厂安全分析报告的标准格式和内容》等。
管理导则的其它部分为研究和试验反应堆、核燃料和物料设备、环境和厂址以及职业保健等。
2.4 美国核管理委员会的技术文件（NUREG）（第四层次）（HAF·J）▲NUREG文件：美国核管理委员会下设的反应堆管理局负责编制的技术文件；▲NUREG/CR文件：委托各种研究机构完成的技术文件。
NUREG文件和NUREG/CR文件属于建议性的参考文件；有时NUREG文件与R.G具有同样的作用：如“NUREG-0800”是《核电厂安全分析报告的标准审查大纲》，这是NRC 对申请者按照“R.G.1.70”《核电厂安全分析报告的标准格式和内容》要求编写的“初步/最终安全分析报告”进行审查的指导性文件。
美国ASME核规范与AP1000编码系统
美国ASME核规范与AP1000编码系统第1章核标准规范的使用依据美国联邦法规10CFR50“Domestic Licnesing of Production and Utilizition Facilities”附录A“设计总则”第2条“针对自然现象防护的设计基准”中要求，安全上重要的构筑物、系统和部件应设计成能承受地震的作用而不丧失其执行安全功能的能力。
这些功能包括：（1）反应堆压力边界的完整性；（2）关停反应堆并将其保持在安全停堆状态的能力；（3）防止事故发生或减轻其后果的能力，这些事故引起的厂外照射可能达到相当于10CFR100规定的照射水平。
10CFR100为“反应堆选址准则”。
为满足此要求，美国核管会（NRC）发布了管理导则RG1.29“抗震设计分类”，介绍了一种可据以确定哪些电厂设施应设计成能承受安全停堆地震的作用的可接受的方法。
10CFR50及其分册50.50a中提出了关于安全上重要的构筑物、系统和部件的设计、制造、安装和试验应达到的质量标准须与其所执行的安全功能的重要性相一致的要求。
为满足此要求，NRC发布了管理导则RG1.26“质量分组和标准”，介绍了一种可据以确定水冷堆安全上重要的容水和容汽部件如何划为不同质量组的可接受的方法。
10CFR50的50.50a中划出了水冷堆安全上重要的部件中属于ASME规范第III篇规范1级部件的部分，它们都是反应堆压力边界的一部分。
这些部件在管理导则RG1.26中均属于质量A组。
安全上重要性较小的含水和含汽部件均定为质量D组。
一般来说，NRC对工业标准（如ASME）采取有条件认可的政策，即根据核安全监管的特殊要求，通过对工业标准的审查，指出那些版本是可以接受，在什么条件下可接受，那些条款是不可接受的，并同时给出相对应的可接受的条款。
ASME规范体系中与核有关的有：ASME BV&P《锅炉压力容器规范》ASME0M《核电厂运行和维修规范》ASME AG-1《核级空气和气体处理系统设计规范》ASME NOG-1《高架吊车建造规则》ASME NQA-1《核设施质量保证大纲要求》ASME NUM-1《悬臂或升降吊车建造规则》ASME N278.1《自动和电动安全相关阀门功能规范标准》ASME N509《核电厂空气净化设备和部件》ASME N510《核气处理系统试验》另外，还有与安装直接相关的下列标准：核电厂建造阶段机械设备和系统的安装、检验和测试用附加质量保证要求（1975年发布，1981年又发布了附录）核电厂建造阶段结构混凝土、结构钢、土质和地基的安装、检验以及测试用附加质量保证要求（1978年发布）核电厂零部件的生姜、安装和运输（1981年发布）关于ASME规范及其补遗的适用版本，应查阅10CFR50§50.55a《规范和标准》。
RG1.061 核电厂扩展设计阻尼值 1973
flitntwr 10721\U.S. ATOMIC ENERGY COMMISSIONREGULATORY GUIDEDIRECTORATE OF REGULATORY STANDARDSREGULATORY GUIDE 1.61DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTSA. INTRODUCTIONCriterion 2, "Design Bases for Protection Against Natural Phenomena," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Production and Utilization Facilities," requires, in part, that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of earthquakes. Proposed Appendix A, "Seismic and Geologic Siting Criteria," to 1OCFR Part 100, "Reactor Site Criteria:' would require, in part, that suitable seismic dynamic analysis, such as a time-history or spectral response .analysis, be performed to demonstrate that the structures, systems, and components important to safety will remain functional in the event of a Safe Shutdown Earthquake (SSE). This guide delineates damping values acceptable to the AEC Regulatory staff to be used in the elastic modal dynamic seismic analysis of Seismic Category IP structures, systems, and components. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.B. DISCUSSIONThe energy dissipation within a structure due to material and structural damping while it is responding to an earthquake depends on a number of factors such as types of joints or connections within the structure, the structural material, and the magnitude of deformations experienced. In a dynamic elastic analysis, this energy dissipation usually Is accounted for by specifying an amount of viscous damping that would result in energy 5Structures, systems, and components of a nuclear power plant that are designated as Seismic Category I are designed to withstand the effects of the Safe Shutdown Earthquake (SSE) and remain functional (see Regulatory Guide 1.29, "Seismic Design Classification").dissipation in the analytical model equivalent to that expected to occur as a result of material and structural damping in the real structure.After reviewing a number of applications for. construction permits and operating licenses and after reviewing pertinent literature including Reference I, the AEC Regulatory staff has determined as acceptable, for interim use, the modal damping values shown in Table I of this guide. These modal damping values should be used for all modes considered in elastic spectral or time-history dynamic analyses. Values are tabulated for the two earthquakes, the Safe Shutdown Earthquake and the Operating Basis Earthquake (or % the Safe Shutdown Earthquake), for which nuclear power plants are required to be designed as specified in proposed Appendix A to 10 CFR Part 100, "Seismic and Geologic Siting Criteria."C. REGULATORY POSITION1. The modal damping values expressed as a percentage of critical damping shown in Table I of this guide should be used for viscous modal damping for all modes considered in an elastic spectral or time-history dynamic seismic analysis of the Seismic Category I structures or components specified in the table. The modal damping values specified in Table I are for use in the dynamic analyses associated with two different magnitudes of earthquakes, the Safe Shutdown Earthquake and theOperating Basis Earthquake (or %t he Safe Shutdown Earthquake). These analyses would be required by proposed Appendix A to 10 CFR Part 100, "Seismic and Geologic Siting Criteria."2. Damping values higher than the ones delineated in Table 1 of this guide may be used in a dynamic seismic analysis if documented test data are provided to support higher values.USAEC REGULATORY GUIDESCe od guide w be obtaied by request Iicamtin the divilsons dlmired to the US. Atomic Energy Commisson. Wshington. D.C. 205M.4Reguhtmy Guides ra Issued to descrbe and muks sieithbi, to she publicAttention: Diector of Regulatory StandardL. Comm•nts end egstflom ftr amuhods ecaptable to the AEC Regultory 8taf of kplementinM qWedfic iWt ofIrnprosyents hI thne- guides am encouraged end ehould be eant to the SemetaryVd Conwidwion' regulations, to deilnete tahniques wmd by th sff in of the Comnmikson. US. Aom: Eney Comitsion. Washington. D.C. 20545, evaluating qecific problems or Postubted ecalm" .or to Provide guldnm toAttantlo- Chief. Public Pro• dings StEff.applicent IGuides em rot substitutes for regulatious and compliance with dmu isn ot uequired. Methodeand solutions diffagnt foron hse aut In The igde am Iasued In t he following un broad divisions: the =uide will be acctable I they provido a E mi. f the ndlgt pPquoRtaetactorthe Isuacor tonthtnnce ofa permit or 1E, e bythe Cconisson. 1. eRascto 6. Pro1dts2. Re- rdi and Test Reactors7. Tranportatin3. Fueskend Materials Facilities L.O ciaetol Health published =ud~ will be eavied perlodlcey. a @pproprimt. so accommnodate4. En~omientul end SitingL.AttutRvewetnronts and ts afetnw inforautios or experience. L Msatralsi and Plant Protection 10. Generalp.3. If the maximum combined stresses due to static, seismic, and other dynamic loading are significantly lower than the yield stress and % y ield stress for SSE and % SSE, respectively, in any structure or component,damping values lower than those specified in Table I of this guide should be used for that structure or component to avoid underestimating the amplitude of vibrations or dynamic stresses.TABLE 1DAMPING VALUES'(Percent of Critical Damping)Operating BasisEarthquake or % Safe Safe Shutdown Structure or Component Shutdown EarthquakeDEarthquakeEquipment and large-diameter piping systems',pipe diameter greater than 12 In. (2)3 Small-diameter piping systems, diameter equalto or less than 12 in. (1)2Welded steel structures (2)4Bolted steel structures (4)7Prestressed concrete structures (2)5Reinforced concrete structures (4)7 'Table 1 is derived from the recommendations given In Reference 1.'Ilithe dynamic analysis of active components as defined in Regulatory Guide 1.48, these valuesshould also be used for SSE.Oln'cludes both material and structural damping. If the piping system consists of onlyone or two spans with little structural demplng& use values for small-diameter piping.REFERENCE1. Newmark, N. M., John A. Blume, and Kanwar K. Kapur,"Design Response Spectra for Nuclear Power Plants," ASCEStructural Engineering Meeting, San Francisco, April 1973.1.61-2。
中国与美国核电厂许可证管理程序的比较
中国与美国核电厂许可证管理程序的比较作者：郁祖盛1 美国核电厂许可证管理程序简介美国核管理委员会(NRC)早期对核电厂的许可证管理，是依据美国联邦法规“10CFRPart50—Domestic Licensing of Production and Utilization Facilities”的规定，实行“建造许可证”和“运行许可证”的“二步法”管理程序，详见图1。
图1 美国联邦法规10CFR50的“二步法”核电厂许可证管理程序为了近一步降低新建核电厂的投资风险和技术风险，美国在1989年颁布了新的联邦法规“10CFR Part52—Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants”，即在颁发“早期厂址许可(ESP)” 和“核电厂设计许可证(DC)” 的基础上，对核电厂的建造和运行许可实行联合运行许可证(COL)制度，即“一步法”管理程序。
DC的管理又可分为2个阶段，即核电厂的设计控制文件(DCD)通过NRC 的安全审评取得“最终设计批准(FD A)”，然后通过公众听证(不涉及安全和技术审查)，即可颁发DC。
核电厂的业主在取得COL后，即可开始核电厂的建造，在建造过程中NRC将根据DC阶段确定的“检验、试验、分析、接受准则(ITAAC)” 实施跟踪监督来核实相关检验、试验和分析结果是否满足验收淮则，如果满足并通过公众听证后，不再需要取得任何许可证件，即可装料、运行。
“一步法”管理程序详见图2。
图2 美国联邦法规10CFR52的“一步法”核电厂许可证管理程序2 中国核电厂许可证管理程序简介根据我国核安全法规《HAF001中华人民共和国民用核设施安全监督管理条例》(1986-10-29国务院发布)和《HAF001/01中华人民共和国民用核设施安全监督管理条例实施细则之一——核电厂安全许可证件的申请和颁发》(1993-12-31国家核安全局发布)的规定，我国对核电厂的许可证管理实行“二步法”管理，其管理程序如图3。
美国核电厂风险评估的安全效益_三_[1]
![美国核电厂风险评估的安全效益_三_[1]](https://img.taocdn.com/s3/m/f9e8b04de518964bcf847c0e.png)
25美国核电厂风险评估的安全效益（三）【美国《核新闻》2003年1月刊报道】 委托监管应用美国核管会（NRC ）在监管过程中积累了大量风险知识，并根据从实施概率风险分析（PRA ）中获得的这些知识对监管作出了诸多改进。
本章将对一些比较重要的风险通报应用进行概要介绍。
ATWS （未能紧急停堆的预期瞬态）规则 ATWS 是反应堆事故保护停堆作用失败后的停堆事件。
这个不太可能发生的事件将引起反应堆系统的高压，同时产生远远超出反应堆停堆散热能力的衰变热，因此反应堆必须停堆并保持在次临界状态。
NRC 在1983年发布了ATWS 规则（10 CFR 50.62），通过以下措施降低ATWS 风险：— 降低预期瞬态发生频率； — 提高反应堆保护停堆作用的可靠性； — 增强在上述预防措施失效时的电厂缓解能力。
在这个规则出台之前，没有通过PRA 获得有关ATWS 堆芯损坏频度（CDF ）的风险，但是NRC 估计，该风险要比目前的风险平均高出10倍。
风险降低的主要原因是降低了预期瞬态发生频率，另一个重要原因是提高了停堆作用的可靠性，这种提高主要是因为增加了对停堆开关的检测和维护频率。
此外，通过使用新的缓解系统，增强了ATWS 的缓解能力。
核电厂的PRA 核实了这种改进。
最初的单个电厂审查（IPE ）模型反映了电厂保护停堆频率和当时已发生的变化。
在11台压水堆（PWR ）机组和11台沸水堆（BWR ）机组中，ATWS 是3个风险最高的始发因素之一。
由于紧急停堆频率的降低（见图3）和新增的变化，目前只有1台PWR 和5台BWR 将ATWS 列为贡献最大的3个风险因素之一。
此外， PRA 使业主/经营者能够根据设备可靠性和燃料性能（这两者对ATWS 风险都重要）随时间的变化，对ATWS 的风险水平实施监控。
全厂断电规则虽然风险通报这个术语直到20世纪90年代中期才出现，但是1988年的全厂断电规则（10 CFR 50.63）清楚表明了在风险通报监管方面的变化。
美核管会即将发布《超设计基准事件缓解规则》
核安全国外核新闻2019．2美核管会即将发布《超设计基准事件缓解规则》【世界核新闻网站2019年1月28日报道】美国核管会（NRC ）近日批准发布根据2011年日本福岛核事故经验教训编制的一份安全规则———《超设计基准事件缓解规则》。
这份规则要求美国商业核电厂储备相应的资源并建立相关程序，以便在发生使厂区丧失所有正常电力供应和应急电源以及反应堆不能向环境安全排出堆芯热量的事故后，冷却反应堆堆芯和乏燃料池并保护安全壳。
此外，还必须配备可用于在严重事件发生后可靠测量乏燃料池液位的设备，以及保护堆芯、安全壳和乏燃料池免受外部威胁所需的“资源”。
大部分美国核电厂必须在《联邦登记册》公布这一规则后的两年零30天内满足规则要求。
核管会表示，这一规则将于2019年春公布。
部分电厂（受核管会2013年3月的安全壳通风令约束的电厂）将拥有三年零30天的时间来满足这些要求。
在福岛核事故发生之后的约1年，即2012年3月，核管会下令立即加强全美商业反应堆的安全。
核管会2019年1月24日表示，《超设计基准事件缓解规则》的适用范围比2013年的安全壳通风令、2012年的缓解策略令或此后颁发的新反应堆许可证中所包含的条件更广泛。
因此，一旦缓解规则要求得到落实，以前发布的多项相关命令以及许可证条件将会终止。
缓解规则还设定了结束对已永久关闭电厂的要求的程序。
核管会表示，核管会及核电厂持证者将在规则制定范围之外继续开展各项后福岛工作，包括分析是否有必要根据特定厂址的最新地震和洪水风险评价结果进行额外的安全改进。
核管会表示，其工作人员已就公众对2016年公布的缓解规则草案的评论意见和建议作出回应。
尽管缓解规则已获得批准，但在核管会委员会的五位委员中，有两位表达了不同意见，称缓解规则未要求持证者为根据最新的地震学和水文学知识重新评估洪水和地震危害做好准备。
美国核能协会（NEI ）表示，核管会批准缓解规则，确认了美国核电厂均“受到很好的保护”。
浅析美国核管会对维修规则进行基准检查的经验
张博平，周晓蕊，李晓洋，等.浅析美国核管会对维修规则进行基准检查的经验［J］.核安全，2019，18（6）：30-36.Zhang Boping，Zhou Xiaorui，Li Xiaoyang，et al.Brief Analysis of the Experience in Benchmarking Maintenance Rules by NRC［J］.Nuclear Safety，2019，18（6）：30-36.浅析美国核管会对维修规则进行基准检查的经验张博平，周晓蕊，李晓洋，张适，郑向阳*（生态环境部核与辐射安全中心，北京100082）摘要：本文简要介绍了维修规则在国际和国内核电厂应用的背景，介绍了美国核管会对美国核电厂维修规则进行基准检查后的经验总结，并对检查范围、构筑物、系统和部件的范围筛选、风险重要度确定、性能指标制定、维修有效性评价、维修风险评价和管理等分别给出检查结果的部分实例，以便于国内维修规则试点单位参考和借鉴。
关键词：维修规则；构筑物；系统和设备；基准检查中图分类号：TM623文章标志码：A文章编号：1672-5360（2019）06-0030-07“基于性能、风险指引型”的维修规则（Maintenance Rule，MR）是由美国核管理委员会（NRC）在20世纪90年代初发布的法规［1］，其目的是推动核电厂对安全相关和部分非安全相关的构筑物、系统和设备（SSC）采取更为有效的维修体系，提高设备的可靠性和可用性，减少因维修不当所导致的对安全系统的挑战。
MR关注维修活动的结果所引起的潜在风险，使核电厂能够将资源更为有效地投入到对安全运行有着重要作用的设备维修与试验中，不仅提高了机组安全水平，也使核电厂的经济效益得到了很大的提高。
除美国，也有一些国家和地区陆续推行了以维修规则为基础的管理体系，包括日本、西班牙等，也有国家在积极引进这套体系，包括韩国、南非等。
我国国家核安全局于2017年8月10日发布了《改进核电厂维修有效性的技术政策（试行）》（国核安发〔2017〕173号）［2］，其基本理念和方法借鉴了维修规则，期望通过政策的实施进一步提高核电厂维修的有效性，加强相关维修活动的风险评价及管理。
核电厂安全重要物项的分级识别概述
0引言
目前在我国核安全法规、标准、核电厂的设计规范中
主要采用确定论方法进行安全物项分级(随着概率风险
分析(PSA)技术应用的不断广泛和深入,基于传统确定 论分级所形成的“特殊处理要求”过于保守，给核电厂的
运维带来很多不必要的负担，限制了核电厂的经济性；但
在某些方面又存在不够保守的情况819。因此，为了使核电
义安全重要功能是指其退化或丧失可能会对核电厂纵
深防御、安全裕量、风险造成重大不利影响的功能，包括
设计基准功能、预防和缓解严重事故的功能$物项(SSC)
的安全重要功能是通过结合风险和传统
的
决策过程来确定的，为了维 其功能有 性
SSC的分级采取相应的监督要求。在传统的确定论方法
中，SSC通常被划分为安全相关和非安全相关两大类，在
邱春辉，顾剑峰，马静娴，宋 强，谭 坤
(中机生产力促进中心，北京100044)
摘 要：论文介绍了国内核安全法规和现有标准中基于确定论方法的核电厂安全相关物项分级方法，以及美 国核电领域基于风险指引的安全重要物项分级和筛选识别方法。通过美国核电厂基于风险指引物项分级的应 用实例，说明开展安全重要物项分级的意义和必要性$ 关键词：物项分级&风险指引&安全重要 中图分类号：TK0) 文献标识码：A doi:10.3969/j.issn.1002-6673.2020.01.017
Review of Categorizing and Identifying Structures, Systems, and Components in NPPS According to Their Safety Significance
QIU Chun-Hui, GU Jian-Feng, MA Jing-Xian, SONG Qiang, TAN Kun "China Productivity Center for Machinery,Beijing 100044,China)
美国核管会关于小型模块堆厂址人口要求的考虑及启示
发 布 了 《 选 址 政 策 工 作 组 报 告 》 ( NUREG [ 2]
础。 在 1980 年的 NRC 授权法案中,国会要求将选
12 月,NRC 在修订的联邦法规“ 反应堆 厂 址 准 则”
沟通是非 常 必 要 的。 同 年,NRC 发 布 了 一 份 重 要
的非轻 水 堆 任 务 的 准 备 》 [ 5] , NRC 工 作 人 员 已 经
求过程中需要关注的问题:( 1) 考虑到总体社会风险并从厂址比选的角度 出 发,建 立 一 个 恰 当 的 反 应 堆 距 人 口 集
中居住区( 或人口中心) 边界的距离是必要的。 此 外,从 纵 深 防 御 考 虑,小 型 堆 厂 址 仍 然 需 要 与 人 口 集 中 居 住 区
保持一个适当的距离;( 2) 基于小型堆选址事故后果及影响范围,建立与大 型 商 用 核 动 力 厂 相 同 社 会 风 险 水 平 的
NRC 发 布 了 《 大 都 会 选 址———历 史 的 视 角 》
( NUREG - 0478)
[ 1]
,为 NRC 工 作 人 员 审 查 人 口 集
中居住区附 近 的 厂 址 提 供 了 一 种 方 法, 也 为 申 请
块堆 设 计 的 潜 在 政 策、 许 可 和 关 键 技 术 问 题 》
5( 识别和解决那些影响非轻水堆核动力 厂 监 管 审
( 列于表 1) ,这些 要 求 基 本 上 是 根 据 NRC 对 大 型
( 1. 生态环境部核与辐射安全中心,北京 102445; 2. 北京市辐射安全研究会,北京 100082)
摘 要:本文介绍了美国核管理委员会( NRC) 关 于 反 应 堆 选 址 过 程 中 与 人 口 因 素 相 关 的 审 管 要 求 和 评 估 准 则,
美国核管会提出日本核事故后加强反应堆安全的建议
美国核管会提出日本核事故后加强反应堆安全的建议王政（译）;伍浩松（校）【摘要】【美国核管会网站2011年7月12日报道】在日本福岛第一核电站于2011年3月因地震和海啸影响而发生严重放射性泄漏事故之后，美国核管会（NRC）组建了一个专门工作组，负责对核管会的各项程序和规章进行系统审查，以确定核管会是否需要进一步加强监管体系，并为核管会改进政策方针提出建议。
该工作组充分意识到，【期刊名称】《国外核新闻》【年(卷),期】2011(000)008【总页数】3页(P14-16)【关键词】美国核管会;反应堆安全;日本;核事故;泄漏事故;监管体系;工作组;放射性【作者】王政（译）;伍浩松（校）【作者单位】不详【正文语种】中文【中图分类】TM623.8【美国核管会网站2011年7月12日报道】在日本福岛第一核电站于2011年3月因地震和海啸影响而发生严重放射性泄漏事故之后，美国核管会（NRC）组建了一个专门工作组，负责对核管会的各项程序和规章进行系统审查，以确定核管会是否需要进一步加强监管体系，并为核管会改进政策方针提出建议。
该工作组充分意识到，发生堆芯熔毁以及放射性物质的不可控释放，即使没有造成一例明显健康危害，也是不可接受的。
工作组还认识到，在未来数十年中，美国各地可能有100多台在役核电机组。
基于上述认识，工作组提出了自己的建议。
在研究此次日本核事故可供借鉴之处时，工作组的主要研究内容包括：针对可引发核事故的自然灾难采取的防护措施、缓解措施以及应急准备。
此次日本核事故是由超过电站设计基准的自然灾害（即海啸）引起的。
作为这项工作的一部分，工作组研究了历史上美国核管会要求采取的针对自然灾害的防护措施，以及核管会怎样应对在美国核电站发生的超设计基准事故。
总体而言，工作组发现，核管会目前采用的监管方法包括：• 通过特别规章或总体设计标准（《联邦法规》10章第50部分（10 CFR Part 50）即《生产和使用设备的国内许可》附录A《核电站总体设计标准》），对设计基准事故的防护和缓解措施提出要求；• 通过特别规章（如电站断电、大型火灾和爆炸），对一些超设计基准事故提出要求；• 一些有关在役反应堆严重事故、战略与导则的自愿性行业倡议。
美国联邦法规10 CFR 50附录Ⅰ的修订方案及对我国辐射防护审管的启示
美国联邦法规10 CFR 50附录Ⅰ的修订方案及对我国辐射防护审管的启示陈晓秋;杨端节【期刊名称】《辐射防护》【年(卷),期】2010()1【摘要】本文详细介绍了美国核管理委员会(NRC)对轻水堆的设计目标基准的修订方案和策略,并在此基础上,考虑到我国核电厂址向内陆发展所致公众照射途径的变化,提出了需要明确核动力厂设计目标值的建议,以及应用现行辐射防护相关标准需要关注的问题:(1)ICRP第103号出版物从以前基于过程的实践和干预的方法发展为基于辐射照射情况性质(计划照射、应急照射和现存照射)的方法,应当注意区分不同的照射情况;(2)ICRP第103号出版物在数值上更新了当量剂量和有效剂量的辐射权重因子和组织权重因子,因此,实施剂量评估所采用的剂量转换因子也需要更新。
【总页数】7页(P1-7)【关键词】轻水堆;设计目标;辐射防护;审管【作者】陈晓秋;杨端节【作者单位】环境保护部核与辐射安全中心,北京100082【正文语种】中文【中图分类】TL7;TS262.6【相关文献】1.美国联邦法规(CFR)中词汇翻译实例分析 [J], 陆红;张维2.国际原子能机构核动力厂辐射防护设计理念对修订《核电厂辐射防护设计》的启示 [J], 付强;陈晓秋;岳会国;杨端节;赵善桂3.NRC修订10 CFR Part 20的技术问题及其启示 [J], 杨端节;陈晓秋4."以人为本"是辐射防护之本--浅议"辐射防护"的某些特点和防护体系的修订 [J], 夏益华5.我国汽车政策法规中的10大问题及根源探析——我国汽车政策标准法规制修订体系也需"放管服"改革 [J], 肖献法因版权原因，仅展示原文概要，查看原文内容请购买。
美国核管会要求电力公司提交核电机组设计基础资料
美国核管会要求电力公司提交核电机组设计基础资料
微亮
【期刊名称】《国外核新闻》
【年(卷),期】1997(000)002
【总页数】1页(P)
【作者】微亮
【作者单位】
【正文语种】中文
【中图分类】TL
【相关文献】
1.美国核管会警告核电机组可能存在设计漏洞 [J], 伍浩松（译）
2.美国核管会收到要求换发ABWR设计合格证的申请 [J], 伍浩松（译）;王海丹（校）
3.日本三菱重工向美国核管会提交美国先进压水堆设计认证申请 [J], 王海丹（译）;张炎（校）
4.韩企向美国核管会提交APR-1400设计认证申请 [J], 伍浩松（译）;王政（校）
5.美国核管会要求核燃料设施业主提交自然灾害应对信息 [J], 伍浩松
因版权原因，仅展示原文概要，查看原文内容请购买。
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美国核管会，10 CFR Part 50, Appendix A，核电厂设计总则Appendix A to Part 50--General Design Criteria for Nuclear Power PlantsTable of Contents∙Introduction∙Definitionso Nuclear Power Unito Loss of Coolant Accidentso Single Failureo Anticipated Operational OccurrencesCRITERIAIntroductionPursuant to the provisions of § 50.34, an application for a construction permit must include the principal design criteria for a proposed facility.The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.These General Design Criteria establish minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the Commission. The General Design Criteria are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units.The development of these General Design Criteria is not yet complete. For example, some of the definitions need further amplification. Also, some of the specific design requirements for structures, systems, and components important to safety have not as yet been suitably defined. Their omission does not relieve any applicant from considering these matters in the design of a specific facility and satisfying the necessary safety requirements. These matters include:(1) Consideration of the need to design against single failures of passive components in fluid systems important to safety. (See Definition of Single Failure.)(2) Consideration of redundancy and diversity requirements for fluid systems important to safety. A "system" could consist of a number of subsystems each of which is separately capable of performing the specified system safety function. The minimum acceptable redundancy and diversity of subsystems and components within a subsystem, and the required interconnection and independence of the subsystems have not yet been developed or defined. (See Criteria 34, 35, 38, 41, and 44.)(3) Consideration of the type, size, and orientation of possible breaks in components of the reactor coolant pressure boundary in determining design requirements to suitably protect against postulatedloss-of-coolant accidents. (See Definition of Loss of Coolant Accidents.)(4) Consideration of the possibility of systematic, nonrandom, concurrent failures of redundant elements in the design of protection systems and reactivity control systems. (See Criteria 22, 24, 26, and 29.)It is expected that the criteria will be augmented and changed from time to time as important new requirements for these and other features are developed.There will be some water-cooled nuclear power plants for which the General Design Criteria are not sufficient and for which additional criteria must be identified and satisfied in the interest of public safety. In particular, it is expected that additional or different criteria will be needed to take into account unusual sites and environmental conditions, and for water-cooled nuclear power units of advanced design. Also, there may be water-cooled nuclear power units for which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General Design Criteria must be identified and justified.Definitions and ExplanationsNuclear power unit. A nuclear power unit means a nuclear power reactor and associated equipment necessary for electric power generation and includes those structures, systems, and components required to provide reasonable assurance the facility can be operated without undue risk to the health and safety of the public.Loss of coolant accidents. Loss of coolant accidents mean those postulated accidents that result from the loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system.1Single failure. A single failure means an occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither (1) a single failure of any active component (assuming passive components function properly) nor (2) a single failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions.2Anticipated operational occurrences. Anticipated operational occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps,tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.CriteriaI. Overall RequirementsCriterion 1--Quality standards and records. Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.Criterion 2--Design bases for protection against natural phenomena. Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.Criterion 3--Fire protection. Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designedto assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.Criterion 4--Environmental and dynamic effects design bases. Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.Criterion 5--Sharing of structures, systems, and components. Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.II. Protection by Multiple Fission Product BarriersCriterion 10--Reactor design. The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.Criterion 11--Reactor inherent protection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.Criterion 12--Suppression of reactor power oscillations. The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.Criterion 13--Instrumentation and control. Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.Criterion 14--Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.Criterion 15--Reactor coolant system design. The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.Criterion 16--Containment design. Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.Criterion 17--Electric power systems. An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.Criterion 18--Inspection and testing of electric power systems. Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.Criterion 19--Control room. A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessaryinstrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses under part 52 of this chapter who do not reference a standard design certification, or holders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.III. Protection and Reactivity Control SystemsCriterion 20--Protection system functions. The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.Criterion 21--Protection system reliability and testability. The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.Criterion 22--Protection system independence. The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other definedbasis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.Criterion 23--Protection system failure modes. The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.Criterion 24--Separation of protection and control systems. The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.Criterion 25--Protection system requirements for reactivity control malfunctions. The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.Criterion 26--Reactivity control system redundancy and capability. Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.Criterion 27--Combined reactivity control systems capability. The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that underpostulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.Criterion 28--Reactivity limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.Criterion 29--Protection against anticipated operational occurrences. The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.IV. Fluid SystemsCriterion 30--Quality of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.Criterion 31--Fracture prevention of reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.Criterion 32--Inspection of reactor coolant pressure boundary. Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity,and (2) an appropriate material surveillance program for the reactor pressure vessel.Criterion 33--Reactor coolant makeup. A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.Criterion 34--Residual heat removal. A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.Criterion 35--Emergency core cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) cladmetal-water reaction is limited to negligible amounts.Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.Criterion 36--Inspection of emergency core cooling system. The emergency core cooling system shall be designed to permit appropriate periodicinspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assure the integrity and capability of the system.Criterion 37--Testing of emergency core cooling system. The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.Criterion 38--Containment heat removal. A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following anyloss-of-coolant accident and maintain them at acceptably low levels.Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.Criterion 39--Inspection of containment heat removal system. The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.Criterion 40--Testing of containment heat removal system. The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to the design as practical the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.。