27.120.10 反应堆工程 标准查询与下载

ASTM E2215-15 评估轻水慢化核电反应堆容器监视舱的标准实施规程

4.1 Neutron radiation effects are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters may be made throughout the service life of the reactor to account for these effects. A surveillance program is used to measure changes in the properties of actual vessel materials due to the irradiation environment. This practice describes the criteria that should be considered in evaluating surveillance program test capsules. 4.2 Prior to the first issue date of this standard, the design of surveillance programs and the testing of surveillance capsules were both covered in a single standard, Practice E185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E185 was revised many times (1966, 1970, 1973, 1979, 1982, 1993 and 1998). Therefore, capsules from surveillance programs that were designed and implemented under early versions of the standard were often tested after substantial changes to the standard had been adopted. For clarity, the standard practice for surveillance programs has been divided into the new Practice E185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. Modifications to the standard test program and supplemental tests are described in Guide E636. 4.3 This practice is intended to cover testing and evaluation of all light-water moderated reactor pressure vessel surveillance capsules. The practice is applicable to testing of capsules from surveillance programs designed and implemented under all previous versions of Practice E185. 4.4 The radiation-induced changes in the properties of the reactor pressure vessel are generally monitored by measuring the index temperatures, the upper-shelf energy and the tensile properties of specimens from the surveillance program capsules. The significance of these radiation-induced changes is described in Practice E185. 4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Guide E636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E1921 or J-integral techniques defined in Test Method E1820. Additionally, hardness testing can be used to supplement standard methods as a means of monitoring the irradiation response of the materials. 4.6 The methodology to be used in the analysis and interpretation of neutron dosimetry data and the determination of neutron fluence is defined in Practice E853. 4.7 Guide E900 describes the b......

Standard Practice for Evaluation of Surveillance Capsules from Light-Water Moderated Nuclear Power Reactor Vessels

ASTM E185-15 轻水慢化核电反应堆容器监督程序设计的标准实施规程

4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel. 4.2 The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guides E482 and E853. 4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects. 1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water small molecular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 18201;×8201;1021 neutrons/m 2 (18201;×8201;1017 n/cm2) at the inside surface of the ferritic steel reactor vessel. 1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.......

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

ASTM C1533-15 热室设备的通用设计考量的标准指南

4.1 The purpose of this guide is to provide general guidelines for the design and operation of hot cell equipment to ensure longevity and reliability throughout the period of service. 4.2 It is intended that this guide record the general conditions and practices that experience has shown is necessary to minimize equipment failures and maximize the effectiveness and utility of hot cell equipment. It is also intended to alert designers to those features that are highly desirable for the selection of equipment that has proven reliable in high radiation environments. 4.3 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 hot cell use. 4.4 This guide is intended to be generic and to apply to a wide range of types and configurations of hot cell equipment. 1.1 Intent: 1.1.1 The intent of this guide is to provide general design and operating considerations for the safe and dependable operation of remotely operated hot cell equipment. Hot cell equipment is hardware used to handle, process, or analyze nuclear or radioactive material in a shielded room. The equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. Therefore, the equipment is operated remotely, either with or without the aid of viewing. 1.1.2 This guide may apply to equipment in other radioactive remotely operated facilities such as suited entry repair areas, canyons or caves, but does not apply to equipment used in commercial power reactors. 1.1.3 This guide does not apply to equipment used in gloveboxes. 1.2 Applicability: 1.2.1 This guide is intended for persons who are tasked with the planning, design, procurement, fabrication, installation, or testing of equipment used in remote hot cell environments. 1.2.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded. 1.2.3 The system of units employed in this standard is the metric unit, also known as SI Units, which are commonly used for International Systems, and defined by IEEE/ASTM SI 10: American National Standard for Use of the International System of Units (SI): The Modern Metric System. 1.3 Caveats: 1.3.1 This guide does not ad......

Standard Guide for General Design Considerations for Hot Cell Equipment

ASTM E900-15 预测在反应堆容器材料中辐射感应变化导致温度漂移的标准指南

4.1 Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause non-ductile behavior in the presence of a flaw. Radiation damage to the reactor vessel is compensated for by adjusting the pressure-temperature limits to higher temperatures as the neutron damage accumulates. The present practice is to base that adjustment on the TTS produced by neutron irradiation as measured at the Charpy V-notch 41-J (30-ft·lbf) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of TTS must be made. 4.1.1 In the absence of surveillance data for a given reactor material (see Practice E185 and E2215), the use of calculative procedures are necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to interpolate or extrapolate the data to obtain a TTS for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes. 4.2 Research has established that certain elements, notably copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn), cause a variation in radiation sensitivity of reactor pressure vessel steels. The importance of other elements, such as silicon (Si), and carbon (C), remains a subject of additional research. Copper, nickel, phosphorus, and manganese are the key chemistry parameters used in developing the calculative procedures described here. 4.3 Only power reactor (PWR and BWR) surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/m2 (E > 1 MeV). Differences in fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been accounted for in these procedures. 1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy V-notch data at 41-J (30-ft·lbf) obtained from surveillance programs conducted in several countries for commercial pressurized (PWR) and boiling (BWR) light-water cooled (LWR) power reactors. An embrittlement correlation has been developed from a statistical analysis of the large surveillance database consisting of radiation-induced TTS and related information compiled and analyzed by Subcommittee E10.02. The details of the database and analysis are described in a separate report (1).2,3, This embrittlement c......

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials

ASTM E185-15e1 轻水冷却核反应堆容器的监督程序设计的标准实施规程

4.1 Predictions of neutron radiation effects on pressure vessel steels are considered in the design of light-water moderated nuclear power reactors. Changes in system operating parameters often are made throughout the service life of the reactor vessel to account for radiation effects. Due to the variability in the behavior of reactor vessel steels, a surveillance program is warranted to monitor changes in the properties of actual vessel materials caused by long-term exposure to the neutron radiation and temperature environment of the reactor vessel. This practice describes the criteria that should be considered in planning and implementing surveillance test programs and points out precautions that should be taken to ensure that: (1) capsule exposures can be related to beltline exposures, (2) materials selected for the surveillance program are samples of those materials most likely to limit the operation of the reactor vessel, and (3) the test specimen types are appropriate for the evaluation of radiation effects on the reactor vessel. 4.2 The methodology to be used in estimation of neutron exposure obtained for reactor vessel surveillance programs is defined in Guides E482 and E853. 4.3 The design of a surveillance program for a given reactor vessel must consider the existing body of data on similar materials in addition to the specific materials used for that reactor vessel. The amount of such data and the similarity of exposure conditions and material characteristics will determine their applicability for predicting radiation effects. 1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. New advanced light-water small modular reactor designs with a nominal design output of 300 MWe or less have not been specifically considered in this practice. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) exceeds 18201;×8201;1021 neutrons/m 2 (18201;×8201;1017 n/cm2) at the inside surface of the ferritic steel reactor vessel. 1.3 This practice does not provide specific procedures for monitoring the radiation induced changes in properties beyond the design life. Practice E2215 addresses changes to the withdrawal schedule during and beyond the design life. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

GSO IEC 61224:2014 核反应堆 热阻探测器 (RTD) 的响应时间 原位测量

This International Standard defines the intended criteria for the choice, conception and use of equipment for the measurement of the response time of resistance temperature detectors (RTD) which are used in the safety and control systems of nuclear reactors. The object of this standard is to describe the techniques which can be used for the in situ measurement of RTD response time. These techniques are recommended only when specific tests performed on loops or physical analysis have shown significant possibilities of drift in the response time.

Nuclear reactors - Response time in resistance temperature detectors (RTD) - In situ measurements

GSO IEC 61031:2014 核电机组正常运行和预期运行事故期间安装的面伽马辐射剂量率监测设备的设计、选址和应用标准

This International Standard applies to the design, location and application of installed equipment for monitoring local gamma radiation dose rates within nuclear power plants during normal operation and anticipated operational occurrences. High range area gamma radiation dose rate monitoring equipment for accident conditions currently addressed by IEC 951-3 is not within the scope of this standard. This standard is intended to be used in conjunction with IEC 532 (second edition)*. NOTE - This standard does not apply to the measurement of neutron dose rate. Additional equipment for neutron monitoring may be required, depending on the plant design, if the neutron dose rate makes a substantial contribution to the total dose equivalent to personnel. This standard provides guidelines for the design principles, the location, the application, the calibration, the operation, and the testing of installed equipment for continuously monitoring local gamma radiation dose rates in nuclear power plants under normal operation conditions and anticipated operational occurrences. These instruments are normally referred to as area radiation monitors. Radiation monitors utilized in area radiation monitoring equipment are dealt with IEC 532. As discussed in that standard, measurement of gamma radiation may be expressed by a number of alternative quantities depending on national regulations. However, for this type of instrument, the most likely quantity to be measured is the absorbed dose in air (Gy), or the ambient dose equivalent (Sv).

Design, location and application criteria for installed area gamma radiation dose rate monitoring equipment for use in nuclear power plants during normal operation and anticipated operational occurrences

GSO IEC 60768:2014 核电机组 重要安全设备 正常和紧急情况下处理电路中放射性的连续内部或并列监测装置

Information regarding the levels of radioactive materials in defined process streams of nuclear power plants is necessary to evaluate plant performance, to provide at an early stage information on possible radioactive releases, and to allow plant operators to take actions to control these releases. This International Standard provides criteria for the design, selection, testing, calibration and functional location of equipment for the monitoring of radioactive substances within plantprocess streams during normal operation conditions and anticipated operational occurrences. IEC 60768 is only applicable to continuous in-line or on-line measurement, i.e. monitors of which the detector measures radioactivity by being positioned in the process stream (i.e. immerged in) or adjacent to the process stream (i.e. viewing straight through a pipe or tank). It does not apply to monitors of which the detector measures a representative proportion of the stream at some remote location (sampling assembly), which are within the scope of IEC 60861. IEC 60768 is only applicable to monitors for normal and incident conditions. Process stream radiation monitoring equipment for accident and post-accident conditions are within the scope of IEC 60951-4.

Nuclear power plants - Instrumentation important to safety - Equipment for continuous in-line or on-line monitoring of radioactivity in process streams for normal and incident conditions

GSO IEC 60980:2014 核电厂安全系统中电气设备实现抗震能力的推荐程序

This standard is applicable to electrical equipment and the instrumentation and control equipment (I & C) of the safety system that is used in nuclear power generating stations including components or equipment of any interface whose failure could adversely affect the perform ance of the safety system. This standard presents acceptable seismic qualification methods and requirements to demonstrate that electrical and I & C equipment can perform their safety-related functions during and after an earthquake. As seismic qualification is only a part of equipment qualification, this standard shall be used in conjunction with IEC 780.

Recommended practices for seismic qualification of electrical equipment of the safety system for nuclear generating stations

EJ/T 20045-2014 钠冷快中子增殖堆设计准则核设计

本标准规定了钠冷快中子增殖堆核设计的基本原则。本标准适用于钠冷快中子增殖堆的核设计。

Design criteria for sodium cooled fast breeder reactor.Nuclear design

EJ/T 20042-2014 钠冷快中子增殖堆设计准则安全设计

本标准规定了钠冷快中子增殖堆安全设计的总体要求和基本原则。本标准适用于钠冷快中子增殖堆的设计,其他液态金属冷却的快中子反应堆可以参照执行。

Design criteria for sodium cooled fast breeder reactor.Safety design

EJ/T 20043-2014 钠冷快中子增殖堆设计准则厂址评价

本标准规定了钠冷快中子增殖堆厂址评价的基本原则和要求。本标准适用于钠冷快中子增殖堆厂址评价、设施运行前及运行三个阶段。

Design criteria for sodium cooled fast breeder reactor.Site evaluation

EJ/T 20046-2014 钠冷快中子增殖堆设计准则热工流体力学设计

本标准规定了钠冷快中子增殖堆热工流体力学设计的基本原则。本标准适用于钠冷快中子增殖堆堆芯的热工流体力学设计。

Design criteria for sodium cooled fast breeder reactor.Thermal-hydraulic design

EJ/T 20044-2014 钠冷快中子增殖堆设计准则反应堆结构总体设计

本标准规定了座池式钠冷快中子增殖堆的反应堆结构总体设计的要求,除堆本体外,还包括堆内换料系统设备和部分堆外换料系统设备。本标准适用于座池式钠冷快中子增殖堆的反应堆结构总体设计。

Design criteria for sodium cooled fast breeder reactor.General design of reactor structure

NB/T 20307-2014 核电厂冷却塔环境影响评价技术规范

本标准规定了核电厂冷却塔环境影响评价应遵循的评价原则、方法和要求。本标准适用于各类新建、扩建内陆及滨海核电厂选址阶段、建造阶段、运行阶段的循环冷却水系统冷却塔环境影响评价相关工作。

Technical code for environmental impact assessment of cooling tower in nuclear power plants

NB/T 20306-2014 核电厂水文地质调查与评价技术规范

本标准规定了核电厂各阶段水文地质调查与评价的工作内容及要求。本标准适用于陆上固定式核电厂、核供热厂的水文地质调查与评价,其他核动力工程也可参照使用。

Technical code for hydrogeological investigation and evaluation of nuclear power plants

NB/T 20278-2014 压水堆核电厂应急给水系统设计准则

本标准规定了压水堆核电厂应急给水系统基本的设计要求,它包括与系统设计直接相关的运行、维修和试验要求,但不包括设备的详细设计要求。本标准适用于二代改进型压水堆核电厂应急给水系统的设计,其它同类型核电厂可参照执行。

Design criteria for emergency feedwater system of pressurized water reactor nuclear power plants

NB/T 20299-2014 核电厂温排水环境影响评价技术规范

本标准规定了核电厂温排水环境影响评价的原则、方法及要求,包括内陆核电厂址和滨海核电厂址。

Technical specification in assessing the environmental impact of thermal discharge for nuclear power plants

NB/T 20285-2014 压水堆核电厂反应堆系统设计总要求

本标准规定了压水堆核电厂反应堆系统设计应满足的总要求。本标准适用于压水堆核电厂反应堆系统的设计。

General requirements of reactor system design for pressurized water reactor nuclear power plants

NB/T 20260-2014 压水堆核电厂余热排出系统设计准则

本标准规定了压水堆核电厂余热排出系统基本的设计要求,它包括与系统设计直接相关运行、维修和试验要求,但不包括设备的详细设计要求。本标准适用于二代改进型压水堆核电厂余热排出系统的设计,其它同类型核电厂可参照执行。

Design criteria for residual heat removal system of pressurized water reactor nuclear power plants