F70 辐射防护与监测综合 标准查询与下载

GOST R 51919-2002 放射性核素电离辐射密封源.泄露测试方法

Radionuclide ionizing radiation sealed sources. Leakage test methods

ASTM E944-02 反应堆监测时中子光谱调节法的应用的标准指南,E 706(IIA)

1.1 This guide covers the analysis and interpretation of the physics dosimetry for Light Water Reactor (LWR) surveillance programs. The main purpose is the application of adjustment methods to determine best estimates of neutron damage exposure parameters and their uncertainties. 1.2 This guide is also applicable to irradiation damage studies in research reactors. 1.3 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 Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E706 (IIA)

ANSI N43.17-2002 利用X射线的个人安全性防护系统的辐射安全

This standard applies to security screening systems that utilize x-radiation and are designed to be used on people. Specifically, it applies to systems used to detect objects carried on the individual being exposed. It provides guidelines specific to the

Radiation Safety for Personnel Security Screening Systems Using X-Rays

DIN 25482-2 Bb.1:2000 电离辐射测量的检测极限和判断域.第2部分:不考虑试样处理影响的分光光度测量法计数测量.说明和示例

Detection limit and decision threshold for ionizing radiation measurements - Part 2: Counting measurements by spectrometry, neglecting the influence of sample treatment; comments and examples

DIN 25482-10:2000 电离辐射测得值的检测极限和置信区间.第10部分:一般应用

The document prepared for general application, is intended to provide a basic rule among other standards of the group of DIN 25482, which are valid only for special measuring tasks. It is also intended to give the user of the standard the possibility to choose an adaquate method of evaluation for a actual measurement problem. It is based on the method of th Bayes-Statistics to take into account also insecure dimensions and influences when using it, which appear not accidentally during repeated or counting measurements.#,,#

Detection limit and decision threshold for ionizing radiation measurements - Part 10: General applications

NF M60-516-3*NF ISO 8529-3:2000 中子参考辐射.第3部分:场所和个人剂量计的校准及中子能量响应和角响应的确定

Reference neutron radiations - Part 3 : calibration of area and personal dosimeters and determination of their response as a function of neutron energy and angle of incidence.

JIS Z 7151:2000 固定源散发.含燃气烟道中颗粒物质浓度和质量流率的测定.人工重量分析法

この規格は，ダクト，煙突，排気管などの閉鎖空間を流れるガス流中の粒子状物質の濃度と質量流量を測定するための手分析 量法について規定したもので，濃度範囲0.005-10 g/m3の測定に適用する。0.050 g/m3以下の濃度では誤差は土10%以上になる(12.及び14.を参照)。

Stationary source emissions -- Determination of concentration and mass flow rate of particulate material in gas-carrying ducts -- Manual gravimetric method

DIN 6802-2:1999 中子剂量学.第2部分:用中子流量计算环境和人员剂量等效值的转换系数和辐射防护装置的修整系数

Neutron dosimetry - Part 2: Conversion coefficients for the calculation of ambient and personal dose equivalent from the neutron fluence, and correction factors for radiation protection devices

BS ISO 15567:1998 放射工作用X-射线(韧性辐射)设施中辐射剂量测定的实施规程

1 This practice covers dosimctric procedures to be followed in facility characterization, process qualification, and routine processing using X-rays (bremsstrahlung) to ensure that the entire product has been treated within an acceptable range of absorbed doses. Other procedures related to facility characterization, process qualification, and routine processing that may influence absorbed dose in the product are also discussed. The establishment of effective or regula-tory dose and X-ray energy limits are not within the scope of this practice. 2 In contrast to monoenergetic gamma rays, the bremsstrahlung energy spectrum extends from low values up to the maximum energy of the electrons incident on the X-ray target (see Section 5 and the Appendix). 3 Dosimetry is only one component of a total quality assurance program for an irradiation facility. Other controls besides dosimetry may be required for specific applications such as medical device sterilization and food preservation (see Sections 8, 9, and 10 and Note 8). Note 1—For guidance in the selection, calibration, and use of specific dosimeters and interpretation of absorbed dose in the product from dose measurements, see the documents listed in 2.1 and 2.2. Note 2—Bremsstrahlung characteristics are similar to gamma rays from radioactive isotopes. See Practice E 1204 for the applications of dosimetry in the characterization and operation of gamma-ray irradia-tion facilities. For information concerning electron beam irradiation technology and dosimetry, see Practice E 1431. 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 appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.

Practice for dosimetry in an X-ray (bremsstrahlung) facility for radiation processing

NF M60-305:1999 核能.核燃料技术废物.通过对放射性核素迁移的无损测量评定放射性核素

Nuclear energy. Nuclear fuel technology-Waste. Assessment of radionuclides by non destructive measurement of radionuclide travers.

ISO 15556:1998 剂量系统和辐射过程的校正和选择指南

Guide for selection and calibration of dosimetry systems for radiation processing

DIN 25482-6 Bb.1:1998 电离辐射测量的监测极限和决策界限.第6部分:考虑试样处理影响和仪器影响给出容差的计数测量.说明和示例

Detection limit and decision threshold for ionizing radiation measurements - Part 6: Counting measurements with allowance for the influence of sample treatment and equipment used; comments and examples

DIN 6802-4:1998 中子剂量学.第4部分.用反射率剂量仪进行单个剂量测量的技术

The document describes the methods of measurements and calibration for albedo dosimeters. These dosimeter are used in personal dosimetry for the determination of the equivalent dose of neutron radiation and photon radiation separately.

Neutron dosimetry - Part 4: Measurement technique for individual dosimetry using albedo dosimeters

SY 6322-1997 油(气)田测井用密封型放射源库安全技术要求

本标准规定了油(气)田测井用密封型放射源库的设计、使用管理的基本安全技术要求。 本标准适用于陆上油(气)田测井甩密封型放射源库的设计及使用管理。本标准不适用于非密封型放射源的存放。

Technological requirements of safety for sealed radioactive sources storehouse used in oil and gas field logging

DIN 25482-5 Bb.1:1997 核辐射测量的探测极限和决策门限.不考虑试样处理影响的高分辨力γ射线光谱计数测量

Detection limit and decision threshold for nuclear radiation measurements - Part 5: Counting measurements by high resolution gamma spectrometry, neglecting the influence of sample treatment; examples and comments

DIN 25482-7:1997 电离辐射测得值的检测极限和置信区间.第7部分:放射材料浓缩过程中过滤器上的计数测量

The document concerns part of nuclear measurement technique, in which events (especially pulses) are counted, caused by nuclear radiation of activity on a filter accumulating radionuclides or groups of radionuclides separated on the filter permits the determination of activity concentration in the throughout streaming agent.

Detection limit and decision threshold for ionizing radiation measurements - Part 7: Counting measurements on filters during accumulation of radioactive materials

ARMY MIL-R-51081 A NOTICE 1-1997 M1A1型钴60伽玛放射源

This notice should be filed in fiont of MIL-R-5108lA(CR), dated 10 December 1986.

RADIOACTIVE SOURCE, COBALT 60, GAMMA, M1A1

ASTM C1174-97 包括地质处理高级核废料在内的废料包装材料长期使用性能的预测的标准实施规程

1.1 This practice covers steps for the development of methods to aid in the prediction of the long-term behavior of waste package materials and waste forms used in the geologic disposal of high-level nuclear waste. 1.1.1 These steps include problem definition, testing, and modelling. 1.1.2 The predictions will be based on models derived from interpretation of data obtained from tests and appropriate analogs. 1.1.3 These tests may include but are not limited to the following: 1.1.3.1 Attribute tests, 1.1.3.2 Characterization tests, 1.1.3.3 Accelerated tests, 1.1.3.4 Service condition tests, 1.1.3.5 Analog tests, and 1.1.3.6 Confirmation tests. 1.2 The purpose of this practice is to provide information to serve as part of the basis for performance assessment of a geologic repository. 1.3 This practice does not cover other methods of making predictions such as use of peer judgments. 1.4 This standard does not purport to address 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 Practice for Prediction of the Long-Term Behavior of Waste Package Materials Including Waste Forms Used in the Geologic Disposal of High-Level Nuclear Waste

ASTM E1005-97 反应堆压力容器监测E706(ⅢA)用辐射监视器的应用和分析的试验方法　　

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 2.1 and in Refs 11, 24-27. The measurement results can be used to define corresponding neutron induced reaction rates which 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 (1-29). 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 (1-10, 12-22). 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 (24-30). 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 (flux density) 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 (24-30). 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 cancelling. The benchmark equivalent flux, for the environment tested, can be calculated from a direct ratio of the measured saturated activities in the two environments and the certified benchmark flux (24-30). 1.2 This method is intended to be used in conjunction with ASTM Guides E706 (IIC) and E844. The following existing or proposed ASTM practices, guides, and methods are also directly involved in the physics-dosimetry evaluation of reactor vessel and support structure surveillance measurements: E 706 (O) Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards E 706 (IA), E853 Analysis and Interpretation of Light-Water Reactor Surveillance Results E 706 (IC), E560 Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results E 706 (ID), E693 Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (DPA) E 706 (IE) Damage Correlation for Reactor Vessel Surveillance E 706 (IF), E185 Practice for Conducting Surveillance Tests for Light-Water Nuclear Power Reactor Vessels E 706 (IG) Surveillance Tests for Nuclear Reactor Support Structures E 706 (IH), E636 Practice for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels E 706 (IIA), E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance E 706 (IIB), E1018 Application of ASTM Evaluated Nuclear Data File (ENDF/A)-Cross Section and Uncertainty File E 706 (IID), E482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance

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

EJ/T 1036-1996 辐射工作场所空气取样的一般规定

General practices for air sampling in radiation workplace