F60 核反应堆综合 标准查询与下载

ASTM D4227-99 核设施安全相关地区混凝土表面涂层涂刷油漆工的资格鉴定

1.1 This practice provides a standard qualifying method for coating applicators to verify their proficiency and ability to attain the required quality for application of specified coatings to concrete surfaces including those in safety-related areas in a nuclear facility. 1.2 Variations or simplifications of the practice set forth herein may be appropriate for special coating work such as maintenance. It is not the intent of this practice to mandate a singular basis for all qualifications. 1.3 Evaluation of the coating applicator being qualified in accordance with this practice, shall be by qualified agents as specified in 4.1. Reports shall be prepared as specified in Section 5, and qualification as specified in Section 6. 1.4 It is the intent of this practice to judge only the ability of the coating applicator to apply specified coatings with the proper tools and equipment. 1.5 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.6 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 Qualification of Coating Applicators for Application of Coatings to Concrete Surfaces

ASTM E2005-99 标准和参考中子场中反应堆剂量测定的基准试验的标准指南

1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results.

Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

ANSI/IEEE 845:1999 核电站、控制室和其他周边区域人-机系统性能评价指南

Provides guidance for evaluating human-system performance related to systems, equipment, and facilities in nuclear power generating stations.

Guide to Evaluation of Man – Machine Performance in Nuclear Power Generating Stations, Control Rooms, and Other Peripheries

ASTM E2006-99 轻水反应堆计算的基准试验的标准指南

1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E 706-IIE1) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactory fluences with a higher degree of confidence. 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 Guide for Benchmark Testing of Light Water Reactor Calculations

ASTM D4228-99 核设施安全相关地区钢表面涂层涂刷油漆工的资格鉴定

1.1 This practice provides a standard qualifying method for coating applicators to verify their proficiency and ability to attain the required quality for application of specified coatings to steel surfaces including those in safety-related areas in a nuclear facility. 1.2 Variations or simplifications of the practice set forth herein may be appropriate for special coating work such as maintenance or qualifications of equipment suppliers shop personnel. It is not the intent of this practice to mandate a singular basis for all qualifications. 1.3 Evaluation of the coating applicator being qualified in accordance with this practice, shall be by qualified agents as specified in 4.1. Reports shall be prepared as specified in Section 5, and certification as specified in Section 6. 1.4 It is the intent of this practice to judge only the ability of the coating applicator to apply specified coatings with the proper tools and equipment. 1.5 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.6 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 Qualification of Coating Applicators for Application of Coatings to Steel Surfaces

ASME OMa CODE Addenda-1999 核电厂的运行和维护规程.附录

Code for operation and maintenance of nuclear power plants; Addenda

ANSI/IEEE 933:1999 核电站可靠性工作计划的定义指南

Provides guidance for the definition of a reliability program at nuclear power generating stations. Emphasizes reliability programs during the operating phase of such stations; however, the general approach applies to all phases of the nuclear power generating station life cycle (e.g., design, construction, start-up, operating, and decommissioning).

Guide for the Definition of Reliability Program Plans for Nuclear Power Generating Stations

IEC 61501:1998 核反应堆仪表 宽范围中子通量比率仪 均方电压法

This International Standard applies to instrument and measurement channels which generate a calculation of the mean square voltage (MSV) of a signal arising from a neutron detector, in order to extract from it information relating to the neutron fluence rate of a nuclear reactor. After calibration, this information can be used to derive the relative power and the time constant, for example expressed in terms of period, doubling time, decades per minute or percent per second. The method used to calculate the mean square voltage of the signal is also known as "fluctuation treatment" or "the Campbell method". Associated with other techniques of measurement, such as pulse rate counting or current measurement, the calculation of the mean square voltage allows the assembly of a series of wide range neutron fluence rate measurements for the simplification of nuclear instrumentation systems in the control of nuclear reactors. This standard describes the principles, the terminology, the characteristics, the requirements and the testing methods related to instrumentation and measurement of the neutron fluence rate using MSV techniques for nuclear reactor control. Typical examples of the application of the MSV techniques are given.

Nuclear reactor instrumentation - Wide range neutron fluence rate meter - Mean square voltage method

IEC 61505:1998 核反应堆仪表 沸水反应堆(BWR) 稳定性监测

This International Standard applies to boiling water reactors (BWR) designed to ensure that thermal-hydraulic oscillations are either not possible, or can be reliably and readily detected and suppressed. Compliance with the criteria can be demonstrated by: a) preventing power oscillations; b) detecting and automatically suppressing power oscillations. Monitoring of the reactor stability state, which is the object of this standard, can support a) and b) by providing information on plant stability characteristics. Monitoring will detect the approach to, and occurrence of, oscillations. The purpose of this standard is to - describe appropriate plant parameters for use in stability monitoring; - define analysis methods for relating time-varying plant information to reactor stability figures of merit, such as the decay ratio; - provide technical guidelines regarding stability monitoring functional and performance requirements. The following items are not covered by this standard: a) control system instability; b) recommendations for implementing a particular solution or a combination of solutions; c) requirements for analytical methods associated with each method; d) requirements for manual operator and automatic actions necessary to suppress power oscillations or reduce decay ratios; e) general safety significance of reactor instability. This standard also does not cover predictor type stability monitors (e.g. frequency domain models which may be used on line to calculate decay ratio for current reactor conditions).

Nuclear reactor instrumentation - Boiling water reactors (BWR) - Stability monitoring

NF M60-306:1998 核能.燃料循环技术.废料.用无源中子计数法测定固体废料包装的放射特性

Nuclear energy. Fuel cycle technology. Waste. Determination of radioactive characteristics of solid waste packages by passive neutron counting.

NF M60-310:1998 核能.燃料循环技术.废料.高放射性固化废料形态长期α辐射稳定性的标准测试方法

Nuclear energy. Fuel cycle technology. Waste. Standard method for testing the long term alpha irradiation stability of solidified high level radioactive waste forms.

NF M60-313:1998 核能.燃料循环技术.废物.索格利特模式化学耐久性试验.玻璃化加工材料的应用

Nuclear energy. Fuel cycle technology. Waste. Soxhlet-mode chemical durability test. Application for vitrification process materials.

ASME OM CODE-1998 核电厂的运行和维护规程

Code for operation and maintenance of nuclear power plants

IEEE 603-1998 核电站安全系统准则

This standard establishes minimum functional design criteria for the power, instrumentation, and control portions of nuclear power generating station safety systems.

Criteria for safety systems for nuclear power generating stations

ANSI/ANS 8.1-1998 外部反应堆裂变物质操作中核临界安全

This standard is applicable to operations with fissionable materials outside nuclear reactors, except for the assembly of these materials under controlled conditions, such as in critical experiments. Generalized basic criteria are presented and limits are specified for some single fissionable units of simple shape containing 233U, 235U, or 239Pu, but not for multiunit arrays. Requirements are stated for establishing the validity and areas of applicability of any calculational method used in assessing nuclear criticality safety.

Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors

ISO 11599:1997 含放射废物水压粘合体的气孔和透气性的测定

This International Standard describes the principles and methodologies of measuring both gas open porosity and permeability of hydraulic binder-based matrices used for immobilization of radioactive waste. The measurements can be carried out by using different apparatus designed and constructed on the basis of a few recommended characteristics. The measurements can be performed on samples prepared in a laboratory or taken from industrial production. Samples can be obtained by moulding or by coring a block.

Determination of gas porosity and gas permeability of hydraulic binders containing embedded radioactive waste

DIN 25493:1997 核设施.防止金属部件表面受到组装工具、密封、装璜、包装材料和隔热材料的损伤

Nuclear facilities - Protection of metallic surfaces of structural parts from damage from assembly aids, gaskets, packings, packaging material and thermal insulating materials

NF M60-780-4:1997 原子能.环境中放射性的测量.生物指示.第4部分:样品制备通用指南

Nuclear energy. Measurement of radioactivity in the environment . Bioindicators. Part 4 : general guide for the preparation of samples.

EJ/T 1034-1996 压水堆核电厂结构设计中在役检查的可达性准则

本标准规定了压水堆核电厂核岛系统、设备、部件及其布置的在役检查（包括役前检查）可达性设计准则。 本标准适用于压水堆核电厂核岛系统、设备、部件及其布置在役检查的设计。

Accessibility criteria for in-service inspections in structural design of pressurized water reactor nuclear power plants

IEC TR3 61510:1996 RBMK核反应堆 仪表和控制改进的建议

Object of this report is to identify those problem areas relating to the safe operation of the RBMK reactors and to make recommendations to improve the safety of those reactors.

RBMK nuclear reactors - Proposals for instrumentation and control improvements