F60 核反应堆综合 标准查询与下载

KS C IEC 60557:2004 核反应堆领域的IEC术语

이 규격은 원자력 계측을 다루는 향후 국제(IEC) 규격에 이용되는 용어의 목록을 상세

IEC terminology in the nuclear reactor field

ANSI/ANS 8.14-2004 反应堆外部核设施中可溶性中子吸收剂的使用

Provides guidance for the use of soluble neutron absorbers for criticality accident control. The standard addresses neutron absorber selection, system design and modifications, safety evaluations, and quality control programs.

Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors

EJ/T 1043-2004 核电厂操纵人员的执照考核

本标准规定了核电厂操纵人员申请执照和换发执照考核活动的相关内容以及知识技能的考核要求。 本标准适用于中华人民共和国核电厂操纵人员申请执照和换发执照的考核工作。

Examintion standards for license of operators of nuclear power plants

IEC 62138:2004 核电厂.设备和控制的安全重要性.表现B类或C类功能的基于计算机系统的软件情况

Provides requirements for the software of computer-based Isystems performing functions of safety category B or C as defined by IEC 61226. Complements IEC 60880 and IEC 60880-2, which provide requirements for the software of computer-based Isystems perfor

Nuclear power plants - Instrumentation and control important for safety - Software aspects for computer-based systems performing category B or C functions

IEEE 577-2004 核电站安全系统的设计和运行中的可靠性分析要求

Provides uniform, minimum acceptable requirements for the performance of reliability analyses for safety-related systems found in nuclear facilities.

Requirement for reliability analysis in the design and operation of safety systems for nuclear facilities

ANSI/ANS 2.26-2004 地震设计用核设施结构系统和元部件的分类

A.This standard provides:(i)criteria and guidelines for selecting an SSC Limit State based on its safety and performance requirements and (ii)criteria for selecting the Seismic Design Category (SDC) for nuclear facility structures, systems, and components (SSCs) for the purpose of designing SSCs to withstand earthquakes using methods specified in ASCE XX. B. The standard outlines the essential facility data and safety analyses necessary to support the seismic design categorization process.

Categorization of Nuclear Facility Structures, Systems, and Components for Seismic Design

ANSI/ANS 14.1-2004 快速脉冲反应堆的运行

This standard is for those involved in the design, operation, and review of fast pulse reactors. It has been formulated in general terms to be applicable to all current fast pulse reactors. This standard does not apply to periodically pulsed reactors or booster assemblies.

Operation of Fast Pulse Reactors

ASME BPVC CCB NC-2004 ASME锅炉和压力容器规程.标准件:CCB NC.核元件

ASME Boiler & Pressure Vessel Code - Code Cases: CCB NC; Nuclear Components

IEEE 323-2003 核电站1E级设备的考核

Describes the basic requirements for qualifying Class 1E equipment and interfaces that are to be used in nuclear power generating stations. The principles, methods and procedures described are intended to be used for qualifying equipment, maintaining and extending qualification, and updating qualification, as required, if the equipment is modified.

Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations

NF M60-771:2003 核能.环境放射性的测量.空气.建筑物中的氡222:筛分和补充检查的方法

Nuclear energy - Measurement of environmental radioactivity - Air - Radon 222 in the buildings : methodologies applies for the screening and for the complementary investigations.

DIN 25435-2:2003 轻水反应堆初级冷却剂回路元件运行中检查.第2部分:磁粉检验和渗透检验

The document is intended to be applied for in-service inspections of surfaces using the magnetic particle and penetrant method for primary coolant circuit components of light water reactors. It specifies inspection methods, requirements relating to the inspection personnel, the equipment, the preparation and performance of the inspection as well as data recording and documentation.#,,#

In-service inspections for primary coolant circuit components of light water reactors - Part 2: Magnetic particle and penetrant inspection

DIN 25435-7:2003 轻水反应堆初级冷却剂回路元件运行中检查.第7部分:放射性试验

The document is intended to be applied for in-service radiographic testing for primary coolant circuit pipes of light water reactors. It specifies inspection methods, requirements relating to the inspection personnel, the equipment, the preparation and performance of the inspection as well as data recording and documentation.#,,#

In-service inspections for primary coolant circuit components of light water reactors - Part 7: Radiographic testing

DIN 25435-6:2003 轻水反应堆初级冷却剂回路元件运行中检查.第6部分:蒸汽发生器加热管涡流试验

In-service inspections for primary coolant circuit components of light water reactors - Part 6: Eddy current testing of steam generator heating tubes

DIN 25435-4:2003 轻水反应堆初级冷却剂回路元件运行中检查.第4部分:目视检查

The document is intended to be applied for in-service inspections of primary coolant circuit components of light water reactors and their internals by visual inspection. It specifies inspection methods, requirements relating to the inspection personnel, the equipment, the preparation and performance of the inspection as well as data recording and documentation.#,,#

In-service inspections for primary coolant circuit components of light water reactors - Part 4: Visual inspection

CSA O322-02-2003 用于地基的压力处理木质材料的认证方法 第2版

CSA Guideline CAN/CSA-QSSO is intended to assist decision-makers in effectively managing all types of risk issues, including injury or damage to health, property, the environment, or something else of value.

Certification method for pressure-treated wood materials intended for foundations Second Edition

BS IEC 61888:2003 核电厂.有关安全的仪器仪表.事故保护定值器的测定和维护

Defines the requirements for assuring that automatic setpoints for nuclear safety system instrumentation are established and maintained within specified limits in nuclear power plants and nuclear reactor facilities.

Nuclear power plants - Instrumentation important to safety - Determination and maintenance of trip setpoints

NF M60-805-1:2003 核能.环境中放射性的测量.水.第1部分:荧光测定法测量水中铀的浓度

Nuclear energy - Measurement of radioactivity in the environment - Water - Part 1 : measurement of the concentration of uranium in water by fluorimetry.

NF M60-805-2:2003 核能.环境中放射性的测量.水.第2部分:感应耦合等离子体原子发射光谱法测量水中铀的浓度

Nuclear energy - Measurement of radioactivity in the environment - Water - Part 2 : measurement of the concentration of uranium in water by inductively coupled plasma atomic emission spectroscopy.

ASTM E1005-03 E706(ⅢA)反应堆压力容器监视用辐射测量监视器的应用和分析的标准试验方法　　

1.1 This method describes general procedures for measuring the specific activities of radioactive nuclides produced in radiometric monitors (RMs) by nuclear reactions induced during surveillance exposures for reactor vessels and support structures. More detailed procedures for individual RMs are provided in separate standards identified in 2.1 and in Refs 11, 24-27. The measurement results can be used to define corresponding neutron induced reaction rates that can in turn be used to characterize the irradiation environment of the reactor vessel and support structure. The principal measurement technique is high resolution gamma-ray spectrometry, although X-ray photon spectrometry and Beta particle counting are used to a lesser degree for specific RMs (1-29).1.1.1 The measurement procedures include corrections for detector background radiation, random and true coincidence summing losses, differences in geometry between calibration source standards and the RMs, self absorption of radiation by the RM, other absorption effects, and radioactive decay corrections (1-10, 12-22).1.1.2 Specific activities are calculated by taking into account the time duration of the count, the elapsed time between start of count and the end of the irradiation, the half life, the mass of the target nuclide in the RM, and the branching intensities of the radiation of interest. Using the appropriate half life and known conditions of the irradiation, the specific activities may be converted into corresponding reaction rates (24-30).1.1.3 Procedures for calculation of reaction rates from the radioactivity measurements and the irradiation power time history are included. A reaction rate can be converted to neutron fluence rate and fluence using the appropriate integral cross section and effective irradiation time values, and, with other reaction rates can be used to define the neutron spectrum through the use of suitable computer programs (24-30). 1.1.4 The use of benchmark neutron fields for calibration of RMs can reduce significantly or eliminate systematic errors since many parameters, and their respective uncertainties, required for calculation of absolute reaction rates are common to both the benchmark and test measurements and therefore are self canceling. The benchmark equivalent fluence rates, for the environment tested, can be calculated from a direct ratio of the measured saturated activities in the two environments and the certified benchmark fluence rate(24-30).1.2 This method is intended to be used in conjunction with ASTM Guide E 844. The following existing or proposed ASTM practices, guides, and methods are also directly involved in the physics-dosimetry evaluation of reactor vessel and support structure surveillance measurements:Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706 (O) E 853 Analysis and Interpretation of Light-Water Reactor Surveillance Results, E 706 (IA)E 560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, E 706 (IC)E 693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E 706 (ID)E 185 Practice for Conducting Surveillance Tests for Light-Water Nuclear Power Reactor Vessels, E 706 (IF)E 1035 Practice for Determining Radiation Exposure for Nuclear Reactor Vessel Support Structures, E 706 (IG)E 636 Practice for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH) E 944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)E 1018 Guide for Application of ASTM Evaluated Cross Section and Data File, E 706 (IIB)E 482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E 706 (IID)E 2005 Guide for the Benchmark Testing of Reactor Vessel ......

Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706(IIIA)

ASME RA-Sa Addenda-2003 核电厂设施的概率风险评估

This Standard sets forth the requirements for probabilistic risk assessments (PRAs) used to support risk-informed decisions for commercial light water reactor nuclear power plants, and prescribes a method for applying these requirements for specific applications.

Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications