F80 核仪器与核探测器综合 标准查询与下载

ASTM E2303-11 辐射加工设施上吸收剂量标记的标准指南

This guide is one of a set of guides and practices that provide recommendations for properly implementing dosimetry in radiation processing. In order to understand and effectively use this and other dosimetry standards, consider first “Practice for Dosimetry in Radiation Processing,” ASTM Practice E2628, which describes the basic requirements that apply when making absorbed dose measurements in accordance with the ASTM E10.01 series of dosimetry standards. In addition, ASTM Practice E2628 provides guidance on the selection of dosimetry systems and directs the user to other standards that provide information on individual dosimetry systems, calibration methods, uncertainty estimation and radiation processing applications. Radiation processing is carried out under fixed path conditions where (a) a process load is automatically moved through the radiation field by mechanical means or (b) a process load is irradiated statically by manually placing product at predetermined positions before the process is started. In both cases the process is controlled in such a manner that the process load position(s) and orientation(s) are reproducible within specified limits. Note 28212;Static irradiation encompasses irradiation of the process load using either manual rotation, no rotation or automated rotation. Some radiation processing facilities that utilize a fixed conveyor path for routine processing may also characterize a region within the radiation field for static radiation processing, sometimes referred to as “Off Carrier” processing. Many radiation processing applications require a minimum absorbed dose (to achieve a desired effect or to fulfill a legal requirement), and a maximum dose that can be tolerated (while the product, material or substance still meets functional specifications or to fulfill a legal requirement). Dose mapping is used to: Characterize the radiation process and assess the reproducibility of absorbed-dose values, which may be used as part of operational qualification and performance qualification. Determine the spatial distribution of absorbed doses and the zone(s) of maximum and minimum absorbed doses throughout a process load, which may consist of an actual or simulated product. Establish the relationship between the dose at a routine monitoring position and the dose within the minimum and maximum dose zones established for a process load. Verify mathematical dose calculation methods. See ASTM Guide E2232. Determine the effect of process interruptions on the distribution of absorbed dose and the magnitude of the minimum and maximum doses. Assess the impact on the distribution of absorbed dose and the magnitude of the minimum and maximum doses resulting from the transition from one process load to another where changes, for example, in product density or product loading pattern may occur.1.1 This document provides guidance in determining absorbed-dose distributions (mapping) in products, materials or substances irradiated in gamma, X-ray (bremsstrahlung) and electron beam facilities. Note 18212;For irradiation of food and the radiation sterilization of health care products, other specific ISO and ISO/ASTM standards containing dose mapping requirements exist. For food irradia......

Standard Guide for Absorbed-Dose Mapping in Radiation Processing Facilities

ASTM E720-11 电子辐射强度测试中测定中子光谱的中子传感器的选择和应用标准指南

Because of the wide variety of materials being used in neutron-activation measurements, this guide is presented with the objective of bringing improved uniformity to the specific field of interest here: hardness testing of electronics primarily in critical assembly reactor environments. Note 28212;Some of the techniques discussed are useful for 14-MeV dosimetry. See Test Method E496 for activation detector materials suitable for 14-MeV neutron effects testing. Note 38212;The materials recommended in this guide are suitable for 252Cf or other weak source effects testing provided the fluence is sufficient to generate countable activities. This guide is organized into two overlapping subjects; the criteria used for sensor selection, and the procedures used to ensure the proper determination of activities for determination of neutron spectra. See Terminology E170 and General Methods E181. Determination of neutron spectra with activation sensor data is discussed in Guides E721 and E944.1.1 This guide covers the selection and use of neutron-activation detector materials to be employed in neutron spectra adjustment techniques used for radiation-hardness testing of electronic semiconductor devices. Sensors are described that have been used at many radiation hardness-testing facilities, and comments are offered in table footnotes concerning the appropriateness of each reaction as judged by its cross-section accuracy, ease of use as a sensor, and by past successful application. This guide also discusses the fluence-uniformity, neutron self-shielding, and fluence-depression corrections that need to be considered in choosing the sensor thickness, the sensor covers, and the sensor locations. These considerations are relevant for the determination of neutron spectra from assemblies such as TRIGA- and Godiva-type reactors and from Californium irradiators. This guide may also be applicable to other broad energy distribution sources up to 20 MeV. Note 18212;For definitions on terminology used in this guide, see Terminology E170. 1.2 This guide also covers the measurement of the gamma-ray or beta-ray emission rates from the activation foils and other sensors as well as the calculation of the absolute specific activities of these foils. The principal measurement technique is high-resolution gamma-ray spectrometry. The activities are used in the determination of the energy-fluence spectrum of the neutron source. See Guide E721. 1.3 Details of measurement and analysis are covered as follows: 1.3.1 Corrections involved in measuring the sensor activities include those for finite sensor size and thickness in the calibration of the gamma-ray detector, for pulse-height analyzer deadtime and pulse-pileup losses, and for background radioactivity. 1.3.2 The primary method for detector calibration that uses secondary standard gamma-ray emitting sources is considered in this guide and in General Methods E181. In addition, an alternative method in wh......

Standard Guide for Selection and Use of Neutron Sensors for Determining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics

ANSI/ASTM E1702:2011 辐射处理过程中加马辐射设备剂量测定操作规程

Practice for Dosimetry in a Gamma Irradiation Facility for Radiation Processing

BS IEC 60462:2010 核仪器.闪烁计数器的光电倍增管.试验程序

Nuclear instrumentation. Photomultiplier tubes for scintillation counting. Test procedures

ISO 8769:2010 对照源.表面污染监测仪的校准.α发射体,β发射体和光子发射体

Reference sources - Calibration of surface contamination monitors - Alpha-, beta- and photon emitters

KS C IEC 61563:2009 辐射防护仪器.测量粮食中γ.辐射放射性核素的比活度用的设备

이 표준은 식품의 감마방출 방사성 핵종의 비방사능 또는 체적 방사능을 측정하기 위한 필드(

Radiation protection instrumentation－Equipment for measuring specific activity of gamma-emitting radionuclides in foodstuffs

ISO/ASTM 51538:2009 乙醇.氯苯剂量测量系统使用实施规程

Practice for use of the ethanol-chlorobenzene dosimetry system

ISO/ASTM 51205:2009 硫酸铈和硫酸高铈剂量测定系统使用的实施规程

Practice for use of a ceric-cerous sulfate dosimetry system

ISO/ASTM 51818:2009 80和300 keV间能量辐射处理用电子束设备剂量测定使用规程

Practice for dosimetry in an electron beam facility for radiation processing at energies between 80 and 300 keV

DIN IEC 60671 Berichtigung 1:2008 核电站.使用仪器和控制设备的安全重要性.监视测试.技术勘误DIN IEC 60671(VDE 0491-100)-2007-12

Nuclear power plants - Instrumentation and control systems important to safety - Surveillance testing (IEC 60671:2007), Corrigendum to DIN IEC 60671 (VDE 0491-100):2007-12

ANSI N43.8-2008 工业致电离辐射计量装置的分类

This standard applies to radiation gauging devices that use sealed radioactive source(s) or machine-generated source(s) for the determination or control of thickness, density, level, interface location, particle size distribution or qualitative or quantitative chemical composition. Establishes a system for classification of the gauging devices based on performance specifications relating to radiation safety.

Classification of Industrial Ionizing Radiation Gauging Devices

ISO 12789-1:2008 参考辐射区域.模拟工作场所中子区域.第1部分:生产特性和方法

This part of ISO 12789 gives guidance for producing and characterizing simulated workplace neutron fields that are to be used for calibrating neutron-measuring devices for radiation protection purposes. Both calculation and spectrometric measurement methods are discussed. Neutron energies in these reference fields range from approximately thermal neutron energies to several hundred GeV. The methods of production and the monitoring techniques for the various types of neutron fields are discussed, and the methods of evaluating and reporting uncertainties for these fields are also given.

Reference radiation fields - Simulated workplace neutron fields - Part 1: Characteristics and methods of production

ASTM E720-08 电子辐射强度试验中中子波谱测定用中子探测器的选择和使用标准指南

Because of the wide variety of materials being used in neutron-activation measurements, this guide is presented with the objective of bringing improved uniformity to the specific field of interest here: hardness testing of electronics primarily in critical assembly reactor environments. NOTE 2 Some of the techniques discussed are useful for 14-MeV dosimetry. See Test Method E 496 for activation detector materials suitable for 14-MeV neutron effects testing. NOTE 3 The materials recommended in this guide are suitable for 252Cf or other weak source effects testing provided the fluence is sufficient to generate countable activities. This guide is organized into two overlapping subjects; the criteria used for sensor selection, and the procedures used to ensure the proper determination of activities for determination of neutron spectra. See Terminology E 170 and General Methods E 181. Determination of neutron spectra with activation sensor data is discussed in Guides E 721 and E 944.1.1 This guide covers the selection and use of neutron-activation detector materials to be employed in neutron spectra adjustment techniques used for radiation-hardness testing of electronic semiconductor devices. Sensors are described that have been used at many radiation hardness-testing facilities, and comments are offered in table footnotes concerning the appropriateness of each reaction as judged by its cross-section accuracy, ease of use as a sensor, and by past successful application. This guide also discusses the fluence-uniformity, neutron self-shielding, and fluence-depression corrections that need to be considered in choosing the sensor thickness, the sensor covers, and the sensor locations. These considerations are relevant for the determination of neutron spectra from assemblies such as TRIGA- and Godiva-type reactors and from Californium irradiators. This guide may also be applicable to other broad energy distribution sources up to 20 MeV. Note 1For definitions on terminology used in this guide, see Terminology E 170.1.2 This guide also covers the measurement of the gamma-ray or beta-ray emission rates from the activation foils and other sensors as well as the calculation of the absolute specific activities of these foils. The principal measurement technique is high-resolution gamma-ray spectrometry. The activities are used in the determination of the energy-fluence spectrum of the neutron source. See Guide E 721.1.3 Details of measurement and analysis are covered as follows:1.3.1 Corrections involved in measuring the sensor activities include those for finite sensor size and thickness in the calibration of the gamma-ray detector, for pulse-height analyzer deadtime and pulse-pileup losses, and for background radioactivity.1.3.2 The primary method for detector calibration that uses secondary standard gamma-ray emitting sources is considered in this guide and in General Methods E 181. In addition, an alternative method in which the sensors are activated in the known spectrum of a benchmark neutron field is discussed in Guide E 1018.1.3.3 A data analysis method is presented which accounts for the following: detector efficiency; background subtraction; irradiation, waiting, and counting times; fission yields and gamma-ray branching ratios; and self-absorption of gamma rays and neutrons in the sensors.1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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 Guide for Selection and Use of Neutron Sensors for Determining Neutron Spectra Employed in Radiation-Hardness Testing of Electronics

ASTM E1894-08 选择脉冲X射线源用的剂量测定系统的标准指南

Flash X-ray facilities provide intense bremsstrahlung radiation environments, usually in a single sub-microsecond pulse, which unfortunately, often fluctuates in amplitude, shape, and spectrum from shot to shot. Therefore, appropriate dosimetry must be fielded on every exposure to characterize the environment, see ICRU Report 34. These intense bremsstrahlung sources have a variety of applications which include the following: Generation of X-ray and gamma-ray environments similar to that from a nuclear weapon burst. Studies of the effects of X rays and gamma rays on materials. Studies of the effects of radiation on electronic devices such as transistors, diodes, and capacitors. Vulnerability and survivability testing of military systems and components. Computer code validation studies. This guide is written to assist the experimenter in selecting the needed dosimetry systems (often in an experiment not all radiation parameters must be measured) for use at pulsed X-ray facilities. This guide also provides a brief summary of the information on how to use each of the dosimetry systems. Other guides (see Section 3) provide more detailed information on selected dosimetry systems in radiation environments and should be consulted after an initial decision is made on the appropriate dosimetry system to use. There are many key parameters which describe a flash X-ray source, such as dose, dose rate, spectrum, pulse width, etc., such that typically no single dosimetry system can measure all the parameters simultaneously. FIG. 1 Range of Available Bremsstrahlung Spectra from Flash X-ray Sources1.1 This guide provides assistance in selecting and using dosimetry systems in flash X-ray experiments. Both dose and dose-rate techniques are described. 1.2 Operating characteristics of flash x-ray sources are given, with emphasis on the spectrum of the photon output. 1.3 Assistance is provided to relate the measured dose to the response of a device under test (DUT). The device is assumed to be a semiconductor electronic part or system.

Standard Guide for Selecting Dosimetry Systems for Application in Pulsed X-Ray Sources

KS A ISO 12789:2007 标准中子射线.模拟工作地点中性磁场的发生方法和特性

이 국제 표준은 방사선 방호 목적을 위한 중성자-측정 장치의 교정을 위해서 사용되는 모의

Reference neutron radiations－Characteristics and methods of production of simulated workplace neutron fields

ISO 21909:2005/Cor 1:2007 无源专用中子剂量计.性能和试验要求.技术勘误1

This standard is Passive personal neutron dosemeters - Performance and test requirements; Technical Corrigendum 1

Passive personal neutron dosemeters - Performance and test requirements; Technical Corrigendum 1

JIS Z 4520:2007 锗γ射线探测器的试验规程

この規格は，ゲルマニウムγ線検出器の製造業者及び使用者にとって重要な性能及び特性の試験方法について規定する。主として高分解能γ線スペクトロメトリーに用いるゲルマニウムγ線検出器の試験方法について規定する。

Test procedures for germanium gamma-ray detectors

ASTM C1455-07 用Υ射线光谱法对特殊核物质含量无损分析的标准试验方法

1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U, or 239Pu remaining as holdup in nuclear facilities. Holdup occurs in all facilities where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources (1, 2, 3, 4).1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements; as well as the laws of physics. The measurement process includes defining measurement uncertainties and is sensitive to the form and distribution of the material, various backgrounds, and interferences. The work includes investigation of material distributions within a facility, which could include potentially large holdup surface areas. Nuclear material held up in pipes, ductwork, gloveboxes, and heavy equipment, is usually distributed in a diffuse and irregular manner. It is difficult to define the measurement geometry, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.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 Nondestructive Assay of Special Nuclear Material Holdup Using Gamma-Ray Spectroscopic Methods

KS A ISO 16797:2006 核能.索格利特式耐化学试验.高放射性废弃物用陶瓷容器的应用

이 규격은 초기 용해율을 측정하는 것에 의하여 물질의 화학적 내구성을 분석하기 위한 속슬렛

Nuclear energy－Soxhlet-mode chemical durability test－Application to vitrified matrixes for high-level radioactive waste

KS I ISO 7935:2006 固定源排放.二氧化硫质量浓度的测定.自动测量方法性能特征

이 규격은 고정 오염원 방출에서 이산화황 함유량의 연속 측정을 위한 자동 측정 시스템에

Stationary source emissions－Determinationof the mass concentration of sulfur dioxide(Performance characteristics of automated measuring methods)