F60 核反应堆综合 标准查询与下载

ASTM E1005-03e1 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　　

Radiometric monitors shall provide a proven passive dosimetry technique for the determination of neutron fluence rate (flux density), fluence, and spectrum in a diverse variety of neutron fields. These data are required to evaluate and estimate probable long-term radiation-induced damage to nuclear reactor structural materials such as the steel used in reactor pressure vessels and their support structures. A number of radiometric monitors, their corresponding neutron activation reactions, and radioactive reaction products and some of the pertinent nuclear parameters of these RMs and products are listed in Table 1. Table 2 provides data (35) on the cumulative and independent fission yields of the important fission monitors. Additional fission product reactions that may provide in situ photo fission information will be added to Table 1 as information is developed and verified (23-29, 36-39). 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 and in Refs , . 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 ().1.1.1 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 (, ).1.1.2 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 (). 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 (). 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().

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

IEEE 383-2003 核电站用1E级电缆质量鉴定和现场拼接

Provides general requirements, direction and methods for qualifying Class 1E electric cables, field splices, factory splices and factory rework for service in nuclear power generating stations

Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations

ASTM C1554-03 热室用物料搬运设备的标准指南

1.1 Intent:1.1.1 This guide covers materials handling equipment used in hot cells (shielded cells) for the processing and handling of nuclear and radioactive materials. The intent of this guide is to aid in the selection and design of materials handling equipment for hot cells in order to minimize equipment failures and maximize the equipment utility.1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, installation, maintenance, repair, replacement, and decontamination and decommissioning of materials handling equipment capable of meeting the stringent demands of operating, dependably and safely, in a hot cell environment where operator visibility is limited due to the radiation exposure hazards.1.1.3 This guide may apply to materials handling equipment in other radioactive remotely operated facilities such as suited entry repair areas and canyons, but does not apply to materials handling equipment used in commercial power reactors.1.1.4 This guide covers mechanical master-slave manipulators and electro-mechanical manipulators, but does not cover electro-hydraulic manipulators.1.2 Applicability:1.2.1 This guide is intended to be applicable to equipment used under one or more of the following conditions:1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment.1.2.1.2 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.1.2.1.3 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, e.g., without shielded viewing windows, periscopes, or a video monitoring system.1.3 User Caveats:1.3.1 This standard is not a substitute for applied engineering skills, proven practices and experience. Its purpose is to provide guidance.1.3.1.1 The guidance set forth in this standard relating to design of equipment is intended only to alert designers and engineers to those features, conditions, and procedures that have been found necessary or highly desirable to the design, selection, operation and maintenance of reliable materials handling equipment for the subject service conditions.1.3.1.2 The guidance set forth results from discoveries of conditions, practices, features, or lack of features that were found to be sources of operational or maintenance problems, or causes of failure.1.3.2 This standard does not supersede federal and/or state regulations, or codes applicable to equipment under any conditions.1.3.3 This standard does not cover design features of the hot cell, e.g., windows, drains, and shield plugs. This standard does not cover pneumatic or hydraulic systems. Refer to Guides C 1533, C 1217, and ANS Design Guides for Radioactive Material Handling Facilities Equipment for information and references to design features of the hot cell and other hot cell equipment.1.3.4 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 Materials Handling Equipment for Hot Cells

ASTM E235-03 核设施或其它有高度可靠性的设施用有护罩的K型和N型热电偶的规范

1.1 This specification covers the requirements for sheathed, Type K and N thermocouples for nuclear service. Depending on size, these thermocouples are normally suitable for operating temperatures to 1652 176;F (900 176;C); special conditions of environment and life expectancy may permit their use at temperatures in excess of 2012 176;F (1100 176;C). This specification was prepared specifically to detail requirements for using this type of sheathed thermocouple in nuclear environments. This specification can be used for sheathed thermocouples which are required for laboratory or general commercial applications where the environmental conditions exceed normal service requirements. The intended use of a sheathed thermocouple in a specific nuclear application will require evaluation by the purchaser of the compatibility of the thermocouple, including the effect of the temperature, atmosphere, and integrated neutron flux on the materials and accuracy of the thermoelements in the proposed application. This specification does not attempt to include all possible specifications, standards, etc., for materials that may be used as sheathing, insulation, and thermocouple wires for sheathed-type construction. The requirements of this specification include only the austenitic stainless steels and other alloys as allowed with Specification E 585/E 585M for sheathing, magnesium oxide or aluminum oxide as insulation, and Type K and N thermocouple wires for thermoelements (see Note 1).1.2 General Design8212;Nominal sizes of the finished thermocouples shall be 0.0400 in. (1.016 mm), 0.0625 in. (1.588 mm), 0.125 in. (3.175 mm), 0.1875 in. (4.763 mm), or 0.250 in. (6.350 mm). Sheath dimensions and tolerances for each nominal size shall be in accordance with . The classes of thermocouples covered by this specification are as follows:1.2.1 Class 1 (grounded)8212;Measuring junction electrically connected to conductive sheaths, and1.2.2 Class 2 (ungrounded)8212;Measuring junctions are electrically isolated from conductive sheaths and from reference ground.1.3 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only.1.4 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.

Specification for Thermocouples, Sheathed, Type K and Type N, for Nuclear or for Other High-Reliability Applications

EJ/T 560-2002 轻水冷却反应堆压力容器辐照监督

Radiation surveillance of pressure vessels for Light-Water Cooled Reactors

EJ/T 712-2002 压水堆核电厂反应堆压力容器及其相关设备安装要求

Installation requirements of reactor pressure vessel and its related equipment in PWR nuclear power plant

NF M60-768:2002 核能.环境放射性的测量.氡222.用累积法评估发散物表面放射性活度的方法

Nuclear energy - Measurement of environmental radioactivity - Radon 222 : methods for estimation of the surfacic activity of exhalation by accumulation method.

CSA O322-02-2002 为了保护木质,压力处理木材认证的程序.第2版

Procedure for Certification of Pressure-Treated Wood Materials for Use in Preserved Wood Foundations Second Edition

Procedure for Certification of Pressure-Treated Wood Materials for Use in Preserved Wood Foundations Second Edition

DIN IEC 61839:2002 核电厂.控制室的设计.功能分析和测定 (IEC 61839:2000)

Nuclear power plants - Design of control-rooms - Functional analysis and assignment (IEC 61839:2000)

IEC TR 62096:2002 核电厂.仪表和控制.确定现代化指南

Support owners of an NPP in the decision-making process and in the preparation for partial or complete modernization of the I&C. Provides a summary of the motivating factors for I&C modernization, the principal options for the elaboration of different sc

Nuclear power plants - Instrumentation and control - Guidance for the decision on modernization

NF C01-121/A1:2002 电工词汇.第121章:电磁学

Electrotechnical Vocabulary - Chapter 121 : electromagnetism.

ANSI/ANS 19.1-2002 反应堆设计计算用核数据集

Identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets that meet these criteria for specific reactor types.

Nuclear Data Sets for Reactor Design Calculations

NF C52-300/A11:2002 反应器

Reactors.

ANSI/ANS 2.23-2002 与地震相关的核电厂反应

Specifies actions that the owner of a of nuclear power plant should take in the event of an earthquake. By using this standard and its companion document, proposed American National Standard for Handling and Initial Evaluation of Records Obtained from Nuclear Power Plant Seismic Instrumentation, BSR/ANS 2.10, the owner can evaluate the need for post-earthquake plant shutdown in a timely manner. Also provides guidelines for determining the condition of components, systems, and structures needed for shutdown and criteria for restart when a nuclear power plant is required to shut down following an earthquake.

Nuclear Plant Response to an Earthquake

ANSI/ANS 19.3.4-2002 核反应堆热能沉积率的测定

Provides criteria for: (1) determination of the energy allocation among the principal particles and photons produced in fission, both prompt and delayed; (2) adoption of appropriate treatment of heavy charged particle and electron slowing down in matter; (3) determination of the spatial energy deposition rates resulting from the interactions of neutrons; (4) calculation of the spatial energy deposition rates resulting from the various interactions of photons with matter; and (5) presentation of the results of such computations, including verification of accuracy and specification of uncertainty. This standard addresses the energy generation and deposition rates for all types of nuclear reactors where the neutron reaction rate distribution and photon and beta emitter distributions are known. Its scope is limited to the reactor core and the thermal and biological shielding.

Determination of Thermal Energy Deposition Rates in Nuclear Reactors

NF M60-610-1:2002 辐射防护.放射生物测定的性能标准.第1部分:一般原则

Radiation protection - Performance criteria for radiobioassay - Part 1 : general principles.

ASTM E900-02 在反应堆容器材料中预测辐射感应变化导致温度的变化的标准指南,E 706(IIF)

1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database:1.1.1 Materials:1.1.1.1 A 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.1.1.1.2 Submerged arc welds, shielded arc welds, and electroslag welds for materials in 1.1.1.1.1.1.2 Copper contents within the range from 0 to 0.50 wt %.1.1.3 Nickel content within the range from 0 to 1.3 wt %.1.1.4 Phosphorus content within the range 0 to 0.025 wt %.1.1.5 Irradiation exposure temperature within the range from 500 to 570176;F (260 to 299176;C).1.1.6 Neutron fluence within the range from 1 x 1016 to 8 x 1019 n/cm2 (E > 1 MeV).1.1.7 Neutron energy spectra within the range expected at the reactor vessel core beltline region of light water cooled reactors and fluence rate within the range from 2 x 108 to 1 x 1012 n/cm2s (E > 1 MeV).1.2 The basis for the method of adjusting the reference temperature is discussed in a separate report.1.3 This guide is Part IIF of Master Matrix E 706 which coordinates several standards used for irradiation surveillance of light-water reactor vessel materials. Methods of determining the applicable fluence for use in this guide are addressed in Master Matrix E 706, Practices E 560 (IC) and Guide E 944 (IIA), and Test Method E 1005 (IIIA). The overall application of these separate guides and practices is described in Practice E 853 (IA).1.4 The values given in customary U.S. units are to be regarded as the standard. The SI values given in parentheses are for information only.1.5 This standard guide does not define how the shift in transition temperature should be used to determine the final adjusted reference temperature. (That would typically include consideration of the initial starting point, the predicted shift, and the uncertainty in the shift estimation method.)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 Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

ASTM E900-02(2007) 在反应堆容器材料中预测辐射感应变化导致温度的变化的标准指南,E 706(IIF)

Operation of commercial power reactors must conform to pressure-temperature limits during heatup and cooldown to prevent over-pressurization at temperatures that might cause nonductile behavior in the presence of a flaw. Radiation damage to the reactor vessel beltline region is compensated for by adjusting the pressure-temperature limits to higher temperature as the neutron damage accumulates. The present practice is to base that adjustment on the increase in transition temperature produced by neutron irradiation as measured at the Charpy V-notch 30-ft·lbf (41-J) energy level. To establish pressure temperature operating limits during the operating life of the plant, a prediction of adjustment in transition temperature must be made. 4.1.1 In the absence of surveillance data for a given reactor (see Practice E 185), the use of calculative procedures will be necessary to make the prediction. Even when credible surveillance data are available, it will usually be necessary to extrapolate the data to obtain an adjustment in transition temperature for a specific time in the plant operating life. The embrittlement correlation presented herein has been developed for those purposes. Research has established that certain elements, notably copper and nickel, cause a variation in radiation sensitivity of steels. The importance of other elements, such as phosphorus (P), remains a subject of additional research. Copper and nickel are the key chemistry parameters used in developing the calculative procedures described here. Only power reactor surveillance data were used in the derivation of these procedures. The measure of fast neutron fluence used in the procedure is n/cm2 (E > 1 MeV). Differences in the neutron fluence rate and neutron energy spectra experienced in power reactors and test reactors have not been applied in these procedures. The manner in which these factors were considered is addressed elsewhere.3 1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database:1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.Submerged arc welds, shielded arc welds, and electroslag welds for materials in .1.1.2 Copper contents within the range from 0 to 0.50 wt %.1.1.3 Nickel content within the range from 0 to 1.3 wt %.1.1.4 Phosphorus content within the range 0 to 0.025 wt %.1.1.5 Irradiation exposure temperature within the range from 500 to 570F (260 to 299C).1.1.6 Neutron fluence within the range from 1 1016 to 8 1019 n/cm2 (E > 1 MeV).1.1.7 Neutron energy spectra within the range expected at the reactor vessel core beltline region of light water cooled reactors and fluence rate within the range from 2 108 to 1 1012 n/cm2s (E > 1 MeV).1.2 The basis for the method of adjusting the ......

Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

ASME RA-S-2002 核电厂设施的概率风险评估

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

Probalistic risk assessment for nuclear power plant applications

ASTM E2215-02 评价轻水核电反应堆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 E 185. Between its provisional adoption in 1961 and its replacement linked to this standard, Practice E 185 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 E 185 that covers the design of new surveillance programs and this standard practice that covers the testing and evaluation of surveillance capsules. A future standard is planned which will recommend procedures for modifying and supplementing existing surveillance programs both in terms of design and testing.4.3 This standard 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 E 185.4.4 The radiation-induced changes in the properties of the vessel are generally monitored by measuring the Charpy transition temperature, the Charpy 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 E 185. The application of this data is the subject of Guide E 900 and other documents listed in Section 2.4.5 Alternative methods exist for testing surveillance capsule materials. Some supplemental and alternative testing methods are available as indicated in Practice E 636. Direct measurement of the fracture toughness is also feasible using the To Reference Temperature method defined in Test Method E 1921 or J-integral techniques defined in Test Method E 1820. Additionally hardness testing can be used to supplement standard methods as a means of monitoring the radiation 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 E 853.4.7 Guide E 900 describes the bases used to evaluate the radiation-induced changes in Charpy transition temperature for reactor vessel materials and provides a methodology for predicting future values.1.1 This practice covers the evaluation of test specimens and dosimetry from light water moderated nuclear power reactor pressure vessel surveillance capsules.1.2 This practice is one of a series of standard practices that outline the surveillance program required for nuclear reactor pressure vessels. The surveillance program monitors the radiation-induced changes in the ferritic steels that comprise the beltline of a light-water moderated nuclear reactor pressure vessel.1.3 This practice along with its companion surveillance program practice, Practice E 185, is intended for application in monitoring the properties of beltline materials in any light-water moderated nuclear reactor.1.4 Modifications to the standard test program and supplemental tests will be described in a separate Standard that is under development to accompany this standard practice and Practice E 185.

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