27.120.20 核电站、安全 标准查询与下载

ASTM D8104-17(2023) 确定涂层鉴定试验数据适用性的标准指南

1.1 The purpose of this guide is to identify evaluation variables that can be used to determine whether existing coating qualification test data (for example, design basis accident or DBA, chemical resistance, fire resistance, thermal conductivity, etc.) meet the respective nuclear power plant qualification requirements or whether requalification is required. Guidance on developing a coating qualification test plan/procedure to qualify a new coating is beyond the scope of this standard. 1.2 This guide is intended for use in new construction and for refurbishing existing coating systems applied to concrete and metal substrates within containment. 1.3 This guide is intended for the use by, or under the supervision of, a person knowledgeable in coating technology and coatings used in CSL I applications, such as a person meeting the requirements of Guide D7108 or equivalent. 1.4 The values stated in inch-pound units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.6 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

Standard Guide for Determining Coating Qualification Test Data Applicability

BS ISO 4233:2023 反应堆技术 核聚变反应堆 核聚变反应堆高温承压部件热氦泄漏检测方法

1   Scope This document specifies the methods and techniques for leak tightness assessment of a metallic component at high temperature by measuring its total leakage rates in a vacuum chamber with a tracer gas leak detector and high-pressure helium gas or the gas mixture flowing out of the component as tracer gas during its thermal and pressure cycles at its operating conditions. The minimum detectable leakage rate can be as low as 10 -10  Pa·m 3 /s, depending on the dimension, external configuration complexity and materials of the component, and is strongly related to the test system and the test conditions. This document is applicable for the hot helium leak test of in -vessel components as per its normal operating conditions in nuclear fusion reactors, which operate at elevated temperatures in an ultra-high vacuum environment down to 10 -6  Pa and with inner flowing-coolant at operating pressure. It is also applicable to the overall leak tightness test of welds in other metallic components and equipment that could be evacuated and pressurized, such as pressurized tanks, pipes and valves in power plants, aerospace and other nuclear reactors.

Reactor technology. Nuclear fusion reactors. Hot helium leak testing method for high temperature pressure-bearing components in nuclear fusion reactors

23/30447567 DC BS EN IEC 63423 核电站 对安全很重要的仪表和控制系统 适用于恶劣环境的电缆组件

BS EN IEC 63423. Nuclear Power Plants. Instrumentation and control systems important to safety. Cable assemblies for Harsh Environment Purposes

T/CNSCPA 004-2023 核用特种设备使用管理程序

为规范核用特种设备使用管理，统一核化工核用特种设备使用管理的工作方法，减少核用特种设备安全监督工作的主观性、随意性，增强科学性、客观性。根据《中华人民共和国特种设备安全法》、《特种设备安全监察条例》、《特种设备目录》、《特种设备使用管理规则》以及民用特种设检验与安全技术监察规程等，制定本标准。

Nuclear special equipment use management procedures

T/CNSCPA 005-2023 核用特种设备安全监督管理程序

为规范核用特种设备安全监督过程，统一核化工核用特种设备安全监督的工作方法，减少核用特种设备安全监督工作的主观性、随意性，增强科学性、客观性。根据《中华人民共和国特种设备安全法》、《特种设备安全监察条例》、《特种设备目录》以及民用特种设检验与安全技术监察规程等，制定本标准。

Procedures for safety supervision and administration of special nuclear equipment

ISO 4233:2023 反应堆技术.核聚变反应堆.核聚变堆中高温承压部件的热氦泄漏试验方法

This document specifies the methods and techniques for leak tightness assessment of a metallic component at high temperature by measuring its total leakage rates in a vacuum chamber with a tracer gas leak detector and high-pressure helium gas or the gas mixture flowing out of the component as tracer gas during its thermal and pressure cycles at its operating conditions. The minimum detectable leakage rate can be as low as 10-10 Pa·m3/s, depending on the dimension, external configuration complexity and materials of the component, and is strongly related to the test system and the test conditions. This document is applicable for the hot helium leak test of in-vessel components as per its normal operating conditions in nuclear fusion reactors, which operate at elevated temperatures in an ultra- high vacuum environment down to 10-6 Pa and with inner flowing-coolant at operating pressure. It is also applicable to the overall leak tightness test of welds in other metallic components and equipment that could be evacuated and pressurized, such as pressurized tanks, pipes and valves in power plants, aerospace and other nuclear reactors.

Reactor technology — Nuclear fusion reactors — Hot helium leak testing method for high temperature pressure-bearing components in nuclear fusion reactors

T/CNSCPA 003-2023 Ⅳ、Ⅴ类放射源及Ⅲ类射线装置项目辐射安全许可审评技术指南

《放射性同位素与射线装置安全防护条例》规定，生产、销售、使用放射性同位素和射线装置的单位，应当取得许可证。按照分级许可的规定，许可证由省、自治区、直辖市人民政府生态环境主管部门审批颁发。2021年生态环境部印发《生态环境部深化“证照分离”改革进一步激发市场主体发展活力实施方案》，进一步明确“经省级政府批准后，使用Ⅳ、Ⅴ类放射源和使用Ⅲ类射线装置的单位的辐射安全许可证，可以由设区的市级生态环境主管部门核发。” 销售、使用Ⅳ、Ⅴ类放射源，生产、销售、使用Ⅲ类射线装置的单位的许可证的审批机关包括各省级及设区的市级生态环境部门，为进一步规范许可申请文件的内容要求，指导地方生态环境部门开展许可技术审评工作，特制定本指南。

Standard Review Plans for Radiation Safety License of Category Ⅳ, Ⅴ Radioactive sources and Category Ⅲ Irradiation Devices

T/CNSCPA 002-2023 核电站风险指引型安全裕度评价中耦合分析方法的技术要求

为指导我国核电站风险指引型安全裕度的评价工作，加强风险指引的安全裕度在核电站设计与运行管理中的应用，确保风险指引型安全裕度评价的质量，充分借鉴国内外实践经验并结合我国实际，根据《中华人民共和国核安全法》、《中华人民共和国民用核设施安全监督管理条例》以及《核动力厂设计安全规定》等，制定本标准。

Technical requirements of coupling analysis method in risk informed safety margin evaluation of nuclear power plants

T/CNSCPA 001-2023 核电站风险指引型安全裕度评价的技术要求

为指导我国核电站风险指引型安全裕度的评价工作，加强风险指引的安全裕度在核电站设计与运行管理中的应用，确保风险指引型安全裕度评价的质量，充分借鉴国内外实践经验并结合我国实际，根据《中华人民共和国核安全法》、《中华人民共和国民用核设施安全监督管理条例》以及《核动力厂设计安全规定》等，制定本标准。

Technical requirements of risk informed safety margin evaluation of nuclear power plants

23/30469793 DC BS IEC 60911 核电站 仪表系统 监测加压轻水反应堆堆芯内充分冷却的测量

BS IEC 60911. Nuclear Power Plants. Instrumentation systems. Measurements for monitoring adequate cooling within the core of pressurized light water reactors

23/30469012 DC BS EN IEC 61225 核电站 仪表、控制和电力系统 静态不间断直流和交流电源系统的要求

BS EN IEC 61225. Nuclear power plants. Instrumentation, control and electrical power systems. Requirements for static uninterruptible DC and AC power supply systems

BS IEC/IEEE 62582-4:2022 核电厂 仪表和控制对安全很重要 电气设备状态监测方法 氧化诱导技术

1   Scope This part of IEC/IEEE 62582 specifies methods for condition monitoring of organic and polymeric materials in instrumentation and control systems using oxidation induction techniques in the detail necessary to produce accurate and reproducible measurements. It includes the requirements for sample preparation, the measurement system and conditions, and the reporting of the measurement results. The different parts of IEC/IEEE 62582 are measurement standards, primarily for use in the management of ageing in initial qualification and after installation. IEC/IEEE 62582‑1 includes requirements for the application of the other parts of the IEC/IEEE 62582 series and some elements which are common to all methods. Information on the role of condition monitoring in the qualification of equipment important to safety is found in IEC/IEEE 60780‑323.

Nuclear power plants. Instrumentation and control important to safety. Electrical equipment condition monitoring methods. Oxidation induction techniques

ASTM E1214-11(2023) 反应堆容器监控用熔化线温度监测器的使用标准指南

1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practices E185 and E2215. 1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties. 1.3 The values stated in SI units are to be regarded as standard. The values given in parentheses are mathematical conversions to inch-pound units that are provided for information only and are not considered standard. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. (See Note 1.) 1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance

BS IEC 62397:2022 核电厂 仪表和控制对安全很重要 电阻温度探测器

1   Scope This document describes the requirements for resistance temperature detectors ( RTDs ) suitable for applications in I&C systems important to safety of nuclear power plants. The requirements of RTDs include design, materials, manufacturing, testing, calibration , procurement, and inspection. RTDs used for safety applications in Nuclear Power Plants can be categorized into direct-immersed and thermowell -mounted RTDs . Furthermore, there are RTDs with specific design which cannot be assigned to the categories mentioned above. However, they are also covered by the requirements stated in this document. This document does not cover the design, material selection, and construction of the thermowell , the guide tube, the extension cable, and the temperature transmitter or resistance bridge which may be associated with the RTD .

Nuclear power plants. Instrumentation and control important to safety. Resistance temperature detectors

BS IEC 60951-1:2022 核设施 对安全很重要的仪表系统 事故和事故后状况的辐射监测 一般要求

1   Scope This part of IEC 60951 provides general guidance on the design principles and performance criteria for equipment to measure radiation and fluid (gaseous effluents or liquids) radioactivity levels at nuclear facilities during and after design basis accidents (DBA) and design extension conditions (DEC), including severe accident (SA). This document is limited to equipment for continuous monitoring of radioactivity in design basis accidents (DBA), design extension conditions (DEC), including severe accident (SA) and post-accident conditions. The purpose of this document is to lay down general requirements and give examples of acceptable methods for equipment for continuous monitoring of radioactivity within the facility during and after design basis accidents (DBA), design extension conditions (DEC), including severe accident (SA) in nuclear facilities. It specifies, for the equipment described above, the general characteristics, general test procedures, radiation, electrical, safety and environmental characteristics and the identification and certification of the equipment. If this equipment is part of a centralized system for continuous radiation monitoring in a nuclear facility, there may be additional requirements from other standards related to this system. Sample extraction and laboratory analysis, which are essential to a complete programme of effluent monitoring, are not within the scope of this document.

Nuclear facilities. Instrumentation systems important to safety. Radiation monitoring for accident and post-accident conditions. General requirements

BS IEC 62705:2022 核设施 仪表和控制对安全很重要 辐射监测系统（RMS）：特征和生命周期

1   Scope This document gives requirements for the lifecycle management of radiation monitoring systems (RMS) and gives guidance on the application of existing IEC standards covering the design and qualification of systems and equipment. The purpose of this document is to lay down requirements for the lifecycle management of RMSs and give application guidance. This document is intended to be consistent with the latest versions of International Standards dealing with radiation monitors , sampling of radioactive materials, instruments calibration , hardware and software design, classification, and qualification. Unless otherwise specified in this document, top level IEC SC 45A standard, IEC 61513, and the second level IEC SC 45A standards apply to RMSs. This document is applicable to RMSs installed in nuclear facilities intended for use during normal operation, anticipated operational occurrences ( AOO ), design basis accidents ( DBA ) and design extension conditions ( DEC ), including severe accidents ( SA ). The technical guidance contained in this document applies to nuclear facilities, although the specific functions of individual facilities are considered during the design and operational lifecycle of RMSs. Laboratory analysis and associated sample extraction, which are essential to a complete programme of effluent monitoring, and investigation for fuel removal are not in the scope of this document.

Nuclear facilities. Instrumentation and control important to safety. Radiation monitoring systems (RMS): Characteristics and lifecycle

ISO 16659-1:2022 核设施通风系统.固体吸附剂碘捕集器的现场效率试验方法.第1部分:一般要求

Ventilation systems for nuclear facilities — In-situ efficiency test methods for iodine traps with solid sorbent — Part 1: General requirements

T/CNS 86-2022 核电厂凝汽器钛管涡流检测技术导则

本文件描述了核电厂凝汽器钛管实施内穿过式涡流检测的方法。

Technical guide for eddy current testing in titanium tube of nuclear power plant condenser

T/CNS 69-2022 高温气冷堆核动力厂蒸汽温度控制系统 闭环试验导则

本文件规定了高温气冷堆核动力厂在调试阶段的蒸汽温度控制系统闭环试验的试验目的、先决条件、试验方法、注意事项、试验数据和报告等的要求。

Guidelines for steam temperature control system closed-loop test in high temperature gas cooled reactor nuclear power plants

T/CNS 60-2022 高温气冷堆核动力厂反应堆首次装料及初始临界试验

本文件描述了高温气冷堆核动力厂在调试试验阶段反应堆首次装料及初始临界试验的专门预防措施、先决条件、初始条件、试验步骤及验收准则等。

Reactor first loading and initial criticality test in high temperature gas cooled reactor nuclear power plants