F74 辐射防护监测与评价 标准查询与下载

ASTM E264-02 通过镍的辐射活化测定快中子反应速率的标准试验方法

1.1 This test method covers procedures for measuring reaction rates by the activation reaction 58Ni(n,p) 58Co.1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.1 MeV and for irradiation times up to about 200 days in the absence of high thermal neutron fluence rates (for longer irradiations, see Practice E 261).1.3 With suitable techniques fission-neutron fluence rates densities above 107 cm2s 1 can be determined.1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261.1.5 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 Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel

ANSI/ASME N509-2002 核电厂空气净化设备和部件

This Standard covers requirements for the design, construction and qualification and acceptance testing of the air-cleaning units and componenets which make up Engineered Safety Features (ESF) and other high efficiency air and gas treatment systems used in nuclear power plants.

Nuclear Power Plant Air Cleaning Units and Components

EJ/T 1131-2001 核电厂环境辐射监测规定

Regulations for environmental radiation monitoring of nuclear power plant

EJ/T 1128-2001 铀矿冶废石、尾矿土质覆盖厚度及降低氡析出率的计算方法

本标准规定了用夏土实测法和理论计算法确定铀矿治废石和尾矿土质覆盖厚度及降低氧析出率的 计算方法。 本标准适用于铀矿冶废石、尾矿等一切含镭-226 的固体废物的土质覆盖层厚度的确定和降低氧析 出率的计算，可在铀矿冶系统及与钠共生的矿治系统废石与尾矿的治理中实施。

Calculation method of covered soil materials thickness and radon emanation rate attenuation for uranium waste rock and mill tailings

ANSI N13.49-2001 放射性测量的操作和文件

Provides minimum requirements for the performance and documentation of radiological surveys. These requirements are incorporated into procedures for specific survey programs, thereby assuring that survey information is complete, uniform, and sufficient f

Performance and Documentation of Radiological Surveys

EJ/T 1117-2000 土壤中镭-226的放射化学分析方法

Radiochemical analysis of radium - 226 in soil

ASTM E706-01 轻水反应堆堆芯压力容器监视标准的标准主模型,E706(0)

1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel''s service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers'' and users'' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users.1.2 This master matrix is intended as a reference and guide to the preparation, revision, and use of standards in the series and for planning and scheduling purposes. This index is to ensure the accomplishment of an objective irrespective of the time required, the number of ASTM committees concerned, or the complexity of the issues involved.1.3 This master matrix standard provides a guide to ASTM standards related to the energy-critical areas that are to be developed under the category of Fission Reactor Development, Section 10, of Guide E584-77 and as discussed in Practice E583-97.1.4 To account for neutron radiation damage in setting pressure-temperature limits and making fracture analyses (see Refs 2-7, 9-14, 21-57, 63, 69-71, 77, 78, 83-104 and Recommended Guide E509), neutron-induced changes in reactor pressure vessel steel fracture toughness must be predicted, then checked by extrapolation of surveillance program data during a vessel''s service life. Uncertainties in the predicting methodology can be significant. Techniques, variables, and uncertainties associated with the physical measurements of PV and support structure steel property changes are not considered in this master matrix, but elsewhere (1, 3, 4, 10-13, 17, 21, 22-27, 32-39, 42, 43, 45, 49-57, 71, 77, 78, 83, 91-104, and Recommended Guide E509). The techniques, variables and uncertainties related to (1) neutron and gamma dosimetry, (2) physics (neutronics and gamma effects), and (3) metallurgical damage correlation procedures and data are addressed in this master matrix (2,34 ). The main variables of concern to (1), (2), and (3) are as follows:1.4.1 Steel chemical composition and microstructure,1.4.2 Steel irradiation temperature,1.4.3 Power plant configurations and dimensions, from the core edge to surveillance positions and into the vessel and cavity walls,1.4.4 Core power distribution,1.4.5 Reactor operating history,1.4.6 Reactor physics computations,1.4.7 Selection of neutron exposure units,1.4.8 Dosimetry measurements,1.4.9 Neutron spectral effects, and1.4.10 Neutron dose rate effects.1.5 A number of potential methods and standards exist for ensuring the adequacy of fracture control of reactor pressure vessel belt lines under normal and accident loads (1-4, 6, 7, 13, 14, 21-28, 29-34, 52-57, 71, 77, 78, 91, 93, Recommended Guide E509, and 2.3 ASME Standards). As older LWR pressure vessels become more highly irradiated, the predictive capability for changes in toughness must improve. Since during a vessel''s service life an increasing amount of information will be available from test reactor and power reactor surveillance programs, better procedures to evaluate and use this information can and must be developed (1-4, 6, 7, 9-15, 17, 21-34, 52-57, 69, ......

Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

EJ/T 287-2000 氚内照射剂量估算与评价方法

Dose estimation and assessment method for internal exposure of tritium

DIN 25483:2000 使用累计固体剂量计的环境监测法

The methods described in the document are intended to long-term integration of local doses caused by photon radiation in case of monitoring population radiation doses outside nuclear installations subject to authorization by  48 German Radiation Protection Ordinance. The document is to be applied to the operating organization of such plants and to measuring points such solid state dosimeter for environmental monitoring. The measuring results give a contribution for controlling whether the dose limits of  44, 45 German Radiation protection ordinance has bees observed. The document recommends a standardized procedure for the evaluation of local doses with a relative global standard-uncertainty of measurement of max. ?25%.#,,#

Methods of environmental monitoring using integrating solid-state dosimeters

ASTM E1249-00(2005) 使用Co-60源的硅电子设备的辐射硬性试验的最小剂量测定错误标准规程

1.1 This practice covers recommended procedures for the use of dosimeters, such as thermoluminescent dosimeters (TLD's), to determine the absorbed dose in a region of interest within an electronic device irradiated using a Co-60 source. Co-60 sources are commonly used for the absorbed dose testing of silicon electronic devices. Note 18212;This absorbed-dose testing is sometimes called "total dose testing" to distinguish it from "dose rate testing."Note 28212;The effects of ionizing radiation on some types of electronic devices may depend on both the absorbed dose and the absorbed dose rate; that is, the effects may be different if the device is irradiated to the same absorbed-dose level at different absorbed-dose rates. Absorbed-dose rate effects are not covered in this practice but should be considered in radiation hardness testing.1.2 The principal potential error for the measurement of absorbed dose in electronic devices arises from non-equilibrium energy deposition effects in the vicinity of material interfaces.1.3 Information is given about absorbed-dose enhancement effects in the vicinity of material interfaces. The sensitivity of such effects to low energy components in the Co-60 photon energy spectrum is emphasized.1.4 A brief description is given of typical Co-60 sources with special emphasis on the presence of low energy components in the photon energy spectrum output from such sources.1.5 Procedures are given for minimizing the low energy components of the photon energy spectrum from Co-60 sources, using filtration. The use of a filter box to achieve such filtration is recommended.1.6 Information is given on absorbed-dose enhancement effects that are dependent on the device orientation with respect to the Co-60 source.1.7 The use of spectrum filtration and appropriate device orientation provides a radiation environment whereby the absorbed dose in the sensitive region of an electronic device can be calculated within defined error limits without detailed knowledge of either the device structure or of the photon energy spectrum of the source, and hence, without knowing the details of the absorbed-dose enhancement effects.1.8 The recommendations of this practice are primarily applicable to piece-part testing of electronic devices. Electronic circuit board and electronic system testing may introduce problems that are not adequately treated by the methods recommended here.1.9This standard does not purport to address all of the safety problems, 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 Minimizing Dosimetry Errors in Radiation Hardness Testing of Silicon Electronic Devices Using Co-60 Sources

JIS Z 4332:1999 Χ射线及Υ射线个人使用的剂量计通则

General requirements for personal X-ray and gamma-ray dosemeters

BS ISO 8529-3:1998 标准中子辐射.第3部分:场所及个人剂量计的校准和作为中子能量作用及入射角的反应的测定

Calibration of area and personal dosimeters and determination of their response as a function of neutron energy and angle of incidence.

Reference neutron radiations - Calibration of area and personal dosimeters and determination of their response as a function of neutron energy and angle of incidence

DIN V 44801-10:1998 α,γ和β射线表面污染测量仪和监视仪.第10部分:α和β射线安装人员污染监视仪

The document relates to contamination waring assemblies for providing monitoring of radioactive contamination on the surface of clothing or skin personnel. The specification is applicable only to that type of equipment where the user takes no action other than to present himself and/or his hands and feet to the detectors over the area to be monitored.

Surface contamination meters and monitors for alpha-, beta- and gamma-radiation - Part 10: Installed personnel surface contamination monitoring assemblies for alpha and beta radiation (IEC 61098:1992, modified)

ASTM E261-98 用放射性技术测定中子的积分通量率、积分通量和中子波谱

1.1 This practice describes the general procedures for the determination of neutron fluence rate, fluence, and energy spectra from the radioactivity that is induced in a detector specimen. 1.2 The practice is directed toward the determination of these quantities in connection with radiation effects on materials. 1.3 For application of these techniques to reactor vessel surveillance, see also Test Methods E 1005. 1.4 This standard does not purport to address all 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. Note 1-Detailed methods for individual detectors are given in the following ASTM test methods: E 262, E 263, E 264, E 265, E 266, E 343, E 393, E 418, E 481, E 523, E 526, E 704, E 705, and E 854.

Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques

EJ/T 1047-1997 尿中氚的分析方法

Analytical determination of tritium in urine

ASTM E262-97 用放射性技术测定热中子反应和积分通量率的试验方法

1.1 The purpose of this method is to define a general procedure for determining an unknown thermal neutron-fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques: 1.1.1 Absolute counting technique using pure cobalt, pure gold, or cobalt-aluminum or gold-aluminum alloy, 1.1.2 Standard foil technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard foil techniques using pure indium, indium-aluminum alloy, and dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-temperature environments are discussed in 8.2. For those circumstances where the use of cadmium as a thermal shield is undesirable because of potential spectrum perturbations or of temperatures above the melting point of cadmium, the method described in Test Method E481 can be used in some cases. Alternatively, gadolinium filters may be used instead of cadmium. For high temperature applications in which aluminum alloys are unsuitable, other alloys such as cobalt-nickel or cobalt-vanadium have been used. 1.3 Table 1 indicates the useful neutron-fluence ranges for each detector material. 1.4 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 This standard may involve hazardous materials, operations and equipment. This standard does not purport to address all of the safety problems 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 Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques

IEEE/ANSI N42.23-1996 放射性检验实验室测量和相关仪表质量保证

Measurement and associated instrumentation quality assurance for radioassay laboratories

EJ 978-1995 铀地质、矿山、选冶厂辐射工作人员个人监测与管理规定

Personal monitoring and management regulations for radiation workers in uranium geology, mines, and dressing and smelting plants

EJ/T 956-1995 水的放射性组分检测取样规程

Sampling procedures for the detection of radioactive components in water

EJ 943-1995 辐射工作人员个人监测管理规定

Regulations on the Administration of Personal Monitoring of Radiation Workers