F60 核反应堆综合 标准查询与下载

DIN IEC 60780:2000 核电厂.安全系统电气设备.合格鉴定

The document represents the best international consensus on the subject of qualification of electrical equipment of safety systems in nuclear power plants. Its application is promoted internationally. In German the qualification generally applies KTA-specifications, but the procedures and requirements describes in this standard are also applicable in Germany.#,,#

Nuclear power plants - Electrical equipment of the safety system - Qualification (IEC 60780:1998)

BS IEC 61468:2000 核电站.堆芯测试设备.自动力中子探测器的特性和试验方法

This International Standard applies to in-core neutron detectors and instrumentation which are designed for purposes important to safety: protection, control and information. It is restricted to characteristics and test methods for self-powered neutron detectors (SPNDs). Self-powered neutron detectors can be used for neutron fluence rate (flux) measurements and spatial power measurements in nuclear reactors. This standard gives requirements, recommendations and guidance on selection of the type and characteristics of SPNDs for different possible applications of SPNDs. For the principles of overall plant and I&C system design and the purpose of neutron fluence rate measurements, reference should be made to general principles of nuclear reactor instrumentation according to IAEA Codes and Safety Guides and IEC 61513.

Nuclear power plants - In-core instrumentation - Characteristics and test methods of self-powered neutron detectors

BS IEC 61504:2000 核电站.对安全致关重要的测试设备和控制系统.对整个工厂范围的辐射监测

This International Standard provides guidance on the design principles and performance criteria for computer-based radiation monitoring systems (RMS). Such systems are provided to integrate the monitoring of plant-wide processes, effluent streams, and area radiation. This standard describes the integration of functions including equipment such as those described in standards IEC 60761-1, IEC 60761-2, IEC 60761-3, IEC 60761-4, IEC 60761-5, IEC 60768, IEC 60910, IEC 60951-1 IEC 60951-2, IEC 60951-3, IEC 60951-4, IEC 60951-5, IEC 61031, and IEC 61250 into a plant-wide digital system. The requirements of system-level components (central computer, subsystem computers, operator consoles, and inter-connections) are discussed. For detection assemblies, processing units and alarm units, this standard contains only the requirements needed to allow connection into the centralized system. The standards referenced above contain the specific requirements for these components. This standard provides criteria for the interface between monitors of different safety classes. This standard integrates data processing, storage, optimization, and correlation of data flow and displays. This standard defines the communication criteria to link distributed radiation monitoring equipment in the plant with an open architecture configuration. This standard does not apply to the design and testing of detection and measurement assemblies and subassemblies except as necessary to define the interface with the plant-wide system. Certain RMS functions, or a complete centralized radiation monitoring system may be entirely implemented with direct-connected analogue/relay technology. This standard does not apply to such functions or systems.

Nuclear power plants - Instrumentation and control systems important to safety - Plant-wide radiation monitoring

BS IEC 61839:2000 核电站.控制室的设计.功能分析和确定

This International Standard specifies functional analysis and assignment procedures (FA and A, sometimes called allocation of functions) for the design of the control-room system for nuclear power plants and gives rules for developing criteria for the assignment of functions. This standard supplements IEC 60964, which applies to the design of the control-room for nuclear power plants. The purpose of this standard is to provide specific requirements for carrying out the functional analysis and assignment required in 3.1 and 3.2 of IEC 60964, and therefore supersedes the guidance given in A.3.1 and A.3.2 of IEC 60964. This standard is applicable to the design of new control-rooms or to backfits (design renewal and design modifications) to existing control-rooms. In the latter case, particular caution is to be exercised to identify areas indirectly affected as well as those directly affected.

Nuclear power plants - Design of control rooms - Functional analysis and assignment

IEC 62118:2000 核反应堆仪表 VVER型加压轻水反应堆(PWR) 冷停过程中堆芯充分冷却的监测

This International Standard applies to pressurized water reactors (PWRs) of VVER design with configurations similar to those shown in figures 1 and 2, and presents the requirements for monitoring adequate cooling within the core during hot and cold shutdown conditions. Annex B provides more information on VVER operating states. Good international practices to be used when designing new or upgrading existing core cooling monitors for VVER systems are summarized in this standard. This standard does not consider the design details of the different VVER technological systems designs, except to the extent that the design affects the monitoring of core cooling. Requirements are given for core cooling monitoring instrumentation to ensure the safe operation of VVERs during abnormal operation and during and after design basis accidents (DBA). Requirements for core cooling monitoring during conditions beyond a DBA, which could be a specific national requirement or consideration, are not covered here. The core cooling monitoring instrumentation has to function under widely different conditions. The circumstances under which this instrumentation needs to function are described. Descriptions of diverse measuring principles and suitable devices are given along with requirements for the following: · operational conditions; · installation; · operator displays; · testing, calibration and maintenance; · equipment qualification; · documentation; · redundancy.

Nuclear reactor instrumentation - Pressurized water reactor (PWR) of VVER design - Monitoring adequate cooling within the core during shutdown

IEC 61839:2000 核电厂 控制室的设计 功能分析和测定

This International Standard specifies functional analysis and assignment procedures (FA and A, sometimes called allocation of functions) for the design of the control-room system for nuclear power plants and gives rules for developing criteria for the assignment of functions. This standard supplements IEC 60964, which applies to the design of the control-room for nuclear power plants. The purpose of this standard is to provide specific requirements for carrying out the functional analysis and assignment required in 3.1 and 3.2 of IEC 60964, and therefore supersedes the guidance given in A.3.1 and A.3.2 of IEC 60964. This standard is applicable to the design of new control-rooms or to backfits (design renewal and design modifications) to existing control-rooms. In the latter case, particular caution is to be exercised to identify areas indirectly affected as well as those directly affected.

Nuclear power plants - Design of control rooms - Functional analysis and assignment

NF M60-416*NF ISO 13463:2000 核能.轻水反应堆MOX燃料生产用核级二氧化钚粉末.产品规范定义协助导则

Nuclear Energy - Nuclear-grade plutonium dioxide powder for fabrication of light water reactor MOX fuel - Guidelines to help in the definition of a product specification.

BS IEC 62117:2000 核反应堆仪表.压力轻水反应堆(PWR).冷停过程中堆芯充分冷却的监测

This International Standard applies to pressurized water reactors (PWRs) with configurations similar to those shown in figures 1 and 2 and presents requirements for the monitoring of adequate cooling within the core during cold shutdown operations. Adequate core cooling can be achieved only by providing sufficient coolant flow to the core to remove the heat. During cold shutdown operations, core cooling is provided by forced circulation with the residual heat removal system (RHRS). However, in certain shutdown operations when water level in the reactor pressure vessel (RPV) is reduced for maintenance operations, forced circulation may be interrupted, and the core may become overheated. It is important that plant operators have reliable information to confirm that the temperature and flow of coolant circulated through the RPV is adequate to remove heat from the core. This information includes a reliable measurement of water level in the RPV outlet piping used for circulating the flow from the core to the RHRS, as well as measurements of temperature and flow of the coolant. An unreliable measurement of water level can result in interruption of coolant flow and overheating of the core, an event that has occurred at several PWRs. Annex A, which describes some of these events, identifies conditions which should be considered in the design of core cooling monitoring instrumentation. Good international practices to be used when designing new or upgrading existing PWR core cooling monitors are summarized in this standard. Requirements are given in this standard for instrumentation to monitor core cooling for safe operation of PWRs during cold shutdown operations when the coolant temperature is below 100 ℃ (212 °F). Requirements for core cooling monitoring during conditions beyond a design basis accident (DBA), which could be a specific national requirement or consideration, are not covered in this standard. The core cooling monitoring instrumentation for cold shutdown functions when the reactor coolant system (RCS) is configured for cold shutdown maintenance or refuelling. The circumstances under which these measurement systems need to function are described in this standard. Descriptions of diverse measuring principles and suitable devices are given together with requirements for the following: - operational conditions; - installation; - operator displays; - testing, calibration and maintenance; - equipment qualification; - documentation.

Nuclear reactor instrumentation - Pressurized light water reactors (PWR) - Monitoring adequate cooling within the core during cold shutdown

IEC 61504:2000 核电厂 安全重要仪表和控制系统 整个工厂范围的辐射监测

This International Standard provides guidance on the design principles and performance criteria for computer-based radiation monitoring systems (RMS). Such systems are provided to integrate the monitoring of plant-wide processes, effluent streams, and area radiation. This standard describes the integration of functions including equipment such as those described in standards IEC 60761-1, IEC 60761-2, IEC 60761-3, IEC 60761-4, IEC 60761-5, IEC 60768, IEC 60910, IEC 60951-1 IEC 60951-2, IEC 60951-3, IEC 60951-4, IEC 60951-5, IEC 61031, and IEC 61250 into a plant-wide digital system. The requirements of system-level components (central computer, subsystem computers, operator consoles, and inter-connections) are discussed. For detection assemblies, processing units and alarm units, this standard contains only the requirements needed to allow connection into the centralized system. The standards referenced above contain the specific requirements for these components. This standard provides criteria for the interface between monitors of different safety classes. This standard integrates data processing, storage, optimization, and correlation of data flow and displays. This standard defines the communication criteria to link distributed radiation monitoring equipment in the plant with an open architecture configuration. This standard does not apply to the design and testing of detection and measurement assemblies and subassemblies except as necessary to define the interface with the plant-wide system. Certain RMS functions, or a complete centralized radiation monitoring system may be entirely implemented with direct-connected analogue/relay technology. This standard does not apply to such functions or systems.

Nuclear power plants - Instrumentation and control systems important to safety - Plant-wide radiation monitoring

IEC 61468:2000 核电厂 堆芯测量仪表 自给能中子探测器的特性和试验方法

This International Standard applies to in-core neutron detectors and instrumentation which are designed for purposes important to safety: protection, control and information. It is restricted to characteristics and test methods for self-powered neutron detectors (SPNDs). Self-powered neutron detectors can be used for neutron fluence rate (flux) measurements and spatial power measurements in nuclear reactors. This standard gives requirements, recommendations and guidance on selection of the type and characteristics of SPNDs for different possible applications of SPNDs. For the principles of overall plant and I&C system design and the purpose of neutron fluence rate measurements, reference should be made to general principles of nuclear reactor instrumentation according to IAEA Codes and Safety Guides and IEC 61513.

Nuclear power plants - In-core instrumentation - Characteristics and test methods of self-powered neutron detectors

ASTM C1217-00(2012) 核材料及放射性材料处理用设备设计的标准指南

Equipment operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns can only be addressed or are most efficiently addressed during one or the other of these stages. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions. This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use. This guide is intended to be generic and to apply to a wide range of equipment types and configurations. The term equipment is used herein in a generic sense. See 3.2.5 for the definition. This service imposes stringent requirements on the quality and the integrity of the equipment, as follows: Leak tightness is required. This implies containment of liquids at all times, and retention of vapors and gases by means of vessel design, or through means of engineered provisions or operational procedures, or both, that ensure the retention, collection, and treatment of vapors and off-gases when the vessel cannot be fabricated or operated with an air-tight vessel configuration. Radioactive materials must be contained. Equipment must be capable of withstanding rigorous chemical cleaning and decontamination procedures. Equipment must be designed and fabricated to remain dimensionally stable throughout its life cycle. Close fabrication tolerances are required to set nozzles and other datum points in known positions. Fabrication materials must be resistant to radiation damage, or materials subject to such damage must be shielded or placed so as to be readily replaceable. Smooth surface finishes are required. Irregularities that hide and retain radioactive particulates or other adherent contamination must be eliminated. Equipment must be capable of being operated virtually unattended, unseen, and trouble-free over long periods. It is assumed that the radiation hazards, combined with the need for confinement and containment, will necessitate a shielded enclosure cell equipped for some degree of remote handling and processing capability in the transuranic materials handling, processing, or machining operations (see 1.2.2). Equipment intended for use in the processing and incorporation of radioactive wastes in host composites or matrices may operate at high temperatures and pressures and may require engineered provisions for the removal of large heat loads under normal and emergency conditions. The chemical corrosion and erosion conditions encountered in these processes tend to be extremely severe, placing emphasis on design for containment integrity. Maintenance records from the plant or from a plant having a similar processing mission may be available for reference. If available and accessible, these records may offer valuable insight with regard to the causes, frequency, and type of failure experienced for the type and class of equipment being designed and engineered. The constraints cited herein are intended to help the engineer establish conditions aimed toward the following: Enhancing radioactive materials containment integrity, Minimizing the loss of in-process materials or the spread of hazardous radioactive contaminants, Minimizing equipment blemishes or faults that promote the adherence or retention of radiation sources, Facilitating the ease and safety of decontamination and maintenance sequences, and Reducing the failure frequency rate for all types and classes of equipment used in this service. Exclusions: In general, this guide is not intended......

Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials

ASME OMb CODE Addenda-2000 核电厂的运行和维护规程.附录

Code for operation and maintenance of nuclear power plants; Addenda

ASTM C1062-00(2014) 核燃料溶解设施的设计, 制造和安装标准指南

4.1 The purpose of this guide is to provide information that will help to ensure that nuclear fuel dissolution facilities are conceived, designed, fabricated, constructed, and installed in an economic and efficient manner. This guide will help facilities meet the intended performance functions, eliminate or minimize the possibility of nuclear criticality and provide for the protection of both the operator personnel and the public at large under normal and abnormal (emergency) operating conditions as well as under credible failure or accident conditions. 1.1 It is the intent of this guide to set forth criteria and procedures for the design, fabrication and installation of nuclear fuel dissolution facilities. This guide applies to and encompasses all processing steps or operations beyond the fuel shearing operation (not covered), up to and including the dissolving accountability vessel. 1.2 Applicability and Exclusions: 1.2.1 Operations—This guide does not cover the operation of nuclear fuel dissolution facilities. Some operating considerations are noted to the extent that these impact upon or influence design. 1.2.1.1 Dissolution Procedures—Fuel compositions, fuel element geometry, and fuel manufacturing methods are subject to continuous change in response to the demands of new reactor designs and requirements. These changes preclude the inclusion of design considerations for dissolvers suitable for the processing of all possible fuel types. This guide will only address equipment associated with dissolution cycles for those fuels that have been used most extensively in reactors as of the time of issue (or revision) of this guide. (See Appendix X1.) 1.2.2 Processes—This guide covers the design, fabrication and installation of nuclear fuel dissolution facilities for fuels of the type currently used in Pressurized Water Reactors (PWR). Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR) and Heavy Water Reactors (HWR) and the fuel dissolution processing technologies discussed herein. However, much of the information and criteria presented may be applicable to the equipment for other dissolution processes such as for enriched uranium-aluminum fuels from typical research reactors, as well as for dissolution processes for some thorium and plutonium-containing fuels and others. The guide does not address equipment design for the dissolution of high burn-up or mixed oxide fuels. 1.2.2.1 This guide does not address special dissolution processes that may require substantially different equipment or pose different hazards than those associated with the fuel types noted above. Examples of precluded cases are electrolytic dissolution and sodium-bonded fuels processing. The guide does not address the design and fabrication of continuous dissolvers. 1.2.3 Ancillary or auxiliary facilities (for example, steam, cooling water, electrical services) are not covered. Cold chemical feed considerations are addressed briefly. 1.2.4 Dissolution Pretreatment—Fuel pretreatment steps incidental to the preparation of spent fuel assemblies for dissolution reprocessing are not covered by this guide. This exclusion applies to thermal treatment steps such as “Voloxidation” to drive off gases prior to dissolution, to mechanical decladding operations or process steps associated with fuel elements disasse......

Standard Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities

IEEE 379-2000 核电站安全系统中单一故障标准的应用

Covers the application of the single-failure criterion to the electrical power, instrumentation, and control portions of nuclear power generating station safety systems.

Application of the single-failure criterion to nuclear power generating station safety systems

ASTM D3843-00(2008) 核装置防护涂层质量保证的标准实施规程

Quality assurance, as covered by this practice, comprises all those planned and systematic actions necessary to provide adequate confidence that safety-related coating work in nuclear facilities as defined in Guide D 5144, will perform satisfactorily in service. It is not practical to impose all the requirements of this practice on certain specific items that require only a small quantity of coating material. The owner, consistent with his formal Quality Assurance Program, may accept affidavits of compliance or certification attesting to the quality of a shop or field coating for such items. If required by licensing commitment; safety-related coatings that are not qualified or for which the quantification basis is indeterminate as defined in Guide D 5144, shall be identified, quantified, and documented. This practice may be incorporated in a project specification by direct reference or may be used to provide guidelines for the quality assurance program for coatings, on the basis of the owner's requirements. Effective use of this practice may also require the incorporation of applicable sections in project specifications for coatings on concrete, steel, equipment, and other related items.1.1 This practice provides a common basis for, and specifically comprises quality assurance requirements applicable to, safety-related protective coating work in Coating Service Level I areas of nuclear facilities as defined in Guide D 5144. 1.2 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 and health practices and determine the applicability of regulatory limitations prior to use.

Standard Practice for Quality Assurance for Protective Coatings Applied to Nuclear Facilities

IEC 61502:1999 核电厂 加压水反应堆 内部结构振动监测

This standard applies to systems used for monitoring the vibratory behaviour of the internal structures of pressurized water reactors (core barrel, thermal shield, upper and lower core support, etc.) and fuel assemblies on the basis of neutron fluctuations observed outside the vessel and vessel vibrations. The main objective of monitoring described in this standard is to detect degradation of internal structures. Primary circuit measurements can be considered together with internal monitoring to provide further information and make it possible to detect degradation of primary circuit structural supports. This standard covers the system characteristics and gives recommendations for monitoring.

Nuclear power plants - Pressurized water reactors - Vibration monitoring of internal structures

NF M60-312:1999 核能.环境放射性空气的测量.利用喷射技术取样的大气氚的活度浓缩的液体闪烁的测定

Nuclear energy. Measurement of environmental radioactivity-Air. Determination by liquid scintillation of the activity concentration of atmospheric tritium sampled by the sparging technique (air through water).

NF M60-780-6:1999 核能.环境中放射性的测量.生物指示剂.第6部分:淡水生物指示剂取样的通用指南

Nuclear energy. Measurement of radioactivity in the environment-Bioindicators. Part 6 : general guide for sampling fresh water biological indicators.

BS ISO 11933-3:1999 密封箱室部件.第3部分:转运装置如简易门,气闸室,双开门用传送系统和废物桶密封连接件

This part of ISO 11933 specifies the designations and characteristics of the various transfer systems which can be mounted on containment enclosures either used alone or placed behind a shielding wall. These transfer systems may also be used directly on shielding walls made of metal (carbon steel or stainless steal) or of concrete. The systems covered by this part of ISO 11933 are — plain doors, — airlock chambers, — double door transfer systems, — leaktight connections for waste drums. Some of these systems, such as plain doors, airlock chambers or double door transfer systems can be used in addition to components defined in ISO 11933-1 or ISO 11933-2 (glove/bag ports, bungs for glove/bag ports, support rings, welded bags, etc.). Large size doors and airlock chambers, used for personnel or large equipment, are not within the scope of this part of ISO 11933.

Components for containment enclosures - Transfer systems such as plain doors, airlock chambers, double door transfer systems, leaktight connections for waste drums

NF C01-821:1999 电工词汇.第821章:铁路信号和安全装置

Electrotechnical Vocabulary. Chapter 821 : signalling and security apparatus for railways.