F60 核反应堆综合 标准查询与下载

ASTM D4286-08 核动力发电机装置用涂覆承包人资格鉴定的标准实施规程

1.1 This practice provides a criteria guide and procedural method to assist utility owners, architects, engineers, constructors, and other selection agencies in determining the overall qualifications of a coating contractor to execute coating work for the primary containment and other safety-related facilities of light-water nuclear power plants. 1.2 The qualification criteria and requirements address the essential basic capability of a contractor to execute nuclear coating work. Obviously, the specific capability to execute those requirements unique to a given project must also be carefully considered. The evaluation procedure contained in this practice is designed to be adaptive to this detailed final qualification process. Variation or simplification of the practice is appropriate for non safety-related areas of nuclear power plants, fossil fueled facilities, and other industrial projects. 1.3 The overall capability of a contractor to successfully execute the varied and complex requirements of nuclear coating work is dependent upon competency in a variety of essential categories. 1.4 The nine evaluation categories described in Sections 3-12 detail the specific data to be provided by the contractor as essential to determining qualification status. In addition, a tenth untitled category has been provided on the evaluation work described in Section 12 for inclusion of other information pertinent to a specific project. 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 and health practices and determine the applicability of regulatory limitations prior to use.

Standard Practice for Determining Coating Contractor Qualifications for Nuclear Powered Electric Generation Facilities

ANSI/IEEE 382:2006 核电厂安全相关功能的鉴定标准

Provides direction for the implementation of the requirements of IEEE Std 323-2003 as they apply to the specific features of safety-related actuator qualification. Establishes criteria for qualification of safety-related actuators, and actuator components, in nuclear power generating stations in order to demonstrate their ability to perform their intended safety functions under all required conditions.

Standard for Qualification of Safety-Related Actuators for Nuclear Power Generating Stations

ANSI/IEEE 338:2007 核电站安全系统周期监督测试标准

Provides criteria for the performance of periodic testing of nuclear power generating station safety systems. The scope of periodic testing consists of functional tests and checks, calibration verification and time response measurements, as required, to verify that the safety system performs its defined safety function.

Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems

ANSI/ANS 15.1-2007 研究反应堆用技术规范的开发

This standard identifies and establishes the content of technical specifications (TS) for research and test reactors. Areas addressed are: Definitions, Safety Limits (SL), Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO), Surveillance Requirements, Design Features, and Administrative Controls. Sufficient detail is incorporated so that applicable specifications can be derived or extracted.

The Development of Technical Specifications for Research Reactors

ANSI/ANS 8.24-2007 核临界安全计算用中子运输法的证实

This standard provides requirements and recommendations for validation, including establishing applicability, of neutron transport calculational methods used in determining critical or subcritical conditions for nuclear criticality safety analyses.

Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations

ASTM E910-07(2013) 反应堆容器监测用氦聚集流监视器应用和分析的标准试验方法, E706 (IIIC)

5.1 The HAFM test method is one of several available passive neutron dosimetry techniques (see, for example, Methods E854 and E1005). This test method can be used in combination with other dosimetry methods, or, if sufficient data are available from different HAFM sensor materials, as an alternative dosimetry test method. The HAFM method yields a direct measurement of total helium production in an irradiated sample. Absolute neutron fluence can then be inferred from this, assuming the appropriate spectrum integrated total helium production cross section. Alternatively, a calibration of the composite neutron detection efficiency for the HAFM method may be obtained by exposure in a benchmark neutron field where the fluence and spectrum averaged cross section are both known (see Matrix E706 IIE). 5.2 HAFMs have the advantage of producing an end product, helium, which is stable, making the HAFM method very attractive for both short-term and long-term fluence measurements without requiring time-dependent corrections for decay. HAFMs are therefore ideal passive, time-integrating fluence monitors. Additionally, the burnout of the daughter product, helium, is negligible. 5.2.1 Many of the HAFM materials can be irradiated in the form of unencapsulated wire segments (see 1.1.2). These segments can easily be fabricated by cutting from a standard inventoried material lot. The advantage is that encapsulation, with its associated costs, is not necessary. In several cases, unencapsulated wires such as Fe, Ni, Al/Co, and Cu, which are already included in the standard radiometric (RM) dosimetry sets (Table 1) can be used for both radiometric and helium accumulation dosimetry. After radiometric counting, the samples are later vaporized for helium measurement.TABLE 1 Neutron Characteristics of Candidate HAFM Materials for Reactor Vessel Surveillance HAFM Sensor Material Principal Helium Producing Reaction Thermal Neutron Cross Section, (b) Fission Neutron Spectrum

Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

ANSI/NIRMA CM1.0-2007 核设备结构管理指南

This standard provides guidelines for the development and implementation of Configuration Management (CM) Programs by organizations operating nuclear facilities. The principle focus is on establishing essential elements of a CM Program and identifying associated industry-acceptable approaches for satisfying the requirements for each element. Its purpose is to be a guidance document not a prescriptive standard. The goal is to ensure the consistency between the design requirements, physical installations and related facility information for nuclear power plants and other nuclear facilities as applicable.

Guidelines for Configuration Management for Nuclear Facilities

ASTM E482-07 反应堆压力容器监视用中子输运法应用的标准指南.E706(IID)

1.1 Need for Neutronics Calculations - An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E 944 and Practice E 853 define appropriate computational procedures.1.2 Methodology Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel.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 requirements prior to use.

Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

EJ/T 1207-2006 核电厂运行绩效评估准则

本标准规定了与核电厂运行相关的十个领域的绩效目标以及如何对实现这些目标的有效性进行评估的准则。 本标准适用于核电厂运行评估组织对核电厂进行运行绩效的评估活动，也可用于核电厂业主自我评估。核电厂的运行和生产管理以及即将投运的核电厂的生产准备活动可参照使用。

Criteria for operational performance assessment of nuclear power plants

EJ/T 1209-2006 核电厂培训领域业绩目标和评估准则

本标准规定了核电厂培训业绩目标和对培训实施专项评估的准则。 本标准适用于对核电厂进行培训领域的同行评估活动，也可用于核电厂对培训领域的自我评估。

Performance objectives and assessment criterion on training area of nuclear power plants

KS C IEC 61468:2006 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

이 규격은 안전에 중요한 목적(보호, 제어 및 정보)을 위해 설계된 노 내 핵 계측 검출기

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

KS C IEC 61250:2006 核反应堆.安全重要仪表和控制系统.冷却剂系统的泄漏检测

이 규격은 경수로의 원자로 냉각재 계통으로부터 누설을 검출하는 데 필요한 계측 요건을

Nuclear reactors－Instrumentation and control systems important for safety－Detection of leakage in coolant systems

ASME BPVC Section 11 Addenda 2006-2006 ASME锅炉和压力容器规范.第11部分:核电站组件在使用中进行检验的原则.补充件2006

ASME Boiler & Pressure Vessel Code - Section 11: Rules for Inservice Inspection of Nuclear Power Plant Components; 2006 Addenda July 1, 2006

ASME Ⅲ-3-2006 核设施部件的结构规则

(a) This Article sets forth the requirements for plan-ning, managing, and conducting Quality Assurance Pro-grams for conlxolling the quality of activities performedunder this Division and the rules governing the evaluationof such Programs prior to the issuance of Certificatesfor construction of Class TC and SC containments. TheQuality Assurance requirements for Material Organiza-tions are provided in WA-3800. Certificate Holders areadvised to consult other regulations for quality assurancerequirements governing activities beyond the scope ofthis Division. (b) N3 and NPT Certificate Holders shall comply withthe Basic Requirements and Supplements of ASME NQA-l-1994 Edition, Quality Assurance Program Requirement for Nuclear Facilities, as modified and sup- plemented in WA-4120(b) and WA-4134.

RULES FOR CONSTRUCTION OF NUCLEAR FACILITY COMPONENTS

DIN IEC 62241:2006 核电站.主要控制室.警报功能和表示(IEC 62241-2004)

Nuclear power plants - Main control room - Alarm functions and presentation (IEC 62241:2004)

ANSI/IEEE 336:2005 核装置的1E级电力、仪表和控制设备的安装、检验及试验指南

Provides considerations for pre-installation, installation, inspection, and testing of Class 1E power, instrumentation, and control equipment and systems of a nuclear facility in the process of installing, inspecting, and testing during new construction, modification and maintenance.

Guide for Installation, Inspection, and Testing for Class 1E Power, Instrumentation, and Control Equipment at Nuclear Facilities

ASME BPVC CCB NC Supplement 8-2006 ASME锅炉和压力容器规程.标准件:CCB NC.核元件.补充件8

ASME Boiler & Pressure Vessel Code - Code Cases: CCB NC; Nuclear Components; Supplement 8

CSA N286.7-99(R2007) 核电站科学计算机程序、分析和设计的质量保证

Quality assurance of scientific computer programs, analysis and design of nuclear power plants

SN/T 1659-2005 进口核电设施检验监管规程

本标准规定了进口核电设施检验监管的基本要求。 本标准适用于进口核电设备的检验监督。

Rules for the inspection and supervision of nuclear power plant facilities for import

DIN 25435-1:2005 轻水反应堆初级冷却剂回路元件运行中检查.第1部分:自动超声检查

In-service inspections for primary coolant circuit components of light water reactors - Part 1: Automated ultrasonic inspection