F70 辐射防护与监测综合 标准查询与下载

ISO 27467:2009 核临界安全.假设临界事故的分析

This International Standard specifies areas that are important to study when analysing potential criticality accidents. NOTE 1 It is important that a criticality accident analysis be performed each time a criticality accident is considered credible, due either to criticality contingencies (double batching, procedural violations, etc.) or to the failure of safety provisions (effectiveness of neutron absorber reduced by fire, etc.). NOTE 2 It is important that the criticality safety specialist be mindful that the process of evaluation developed in this International Standard does not address the unforeseen, since any actually occurring criticality accident will probably result from a scenario not envisioned or from failure to comply with prevailing regulations. This International Standard does not address detailed administrative measures, for which the responsibility lies with the public authorities, nor does it deal with criteria used to justify the accident criticality analysis of a nuclear facility. This International Standard does not apply to nuclear power plants.

Nuclear criticality safety - Analysis of a postulated criticality accident

GOST R 53370-2009 辐射防护房间和类似包厢的房间用可移动防辐射聚合物涂层.辐射情况改进.技术要求

Removable protective polymeric covering for radiation protective rooms and box-like premises. Improvement of radioactivity situation. Technological requirements

IEC 62303:2008 辐射防护仪器.空气中氘的监测用设备

Radiation protection instrumentation - Equipment for monitoring of airborne tritium

GJB 6385-2008 军用核材料固定场所实体屏障技术要求

本标准规定了在生产、使用和储存军用核材料的固定场所中，实体屏障的技术要求。 本标准适用于中华人民共和国境内军用核材料许可证持有单位。在设计、建造、验收、使用和维护实体屏障时，这些单位应以本标准为基本依据。其他固定场所实体屏障的构建也可参照本标准执行。

Specification of physical barriers at the fixed sites for military nuclear material

GJB 6386-2008 军用核材料保卫控制中心设计要求

本标准规定了生产、使用、储存军用核材料的固定场所中，保卫控制中心和保卫值班室的设计要求。 本标准适用于中华人民共和国境内军用核材料许可证持有单位保卫控制中心和保卫值班室的设计、建造和验收。

Design criteria of central alarm station for military nuclear material

ANSI N13.54-2008 核医学中胎儿辐射剂量的计算

Radiology and Nuclear Medicine physicians, health physicists, medical physicists, radiation safety officers, regulators and any others with the responsibility for estimating fetal dose should use the methods and numerical estimates shown in the sections

Fetal Radiation Dose Calculations in Nuclear Medicine

ASTM E1006-08 试验反应堆的物理放射量测定结果的分析和解释的标准实施规程,E 706(II)

The mechanical properties of steels and other metals are altered by exposure to neutron radiation. These property changes are assumed to be a function of chemical composition, metallurgical condition, temperature, fluence (perhaps also fluence rate), and neutron spectrum. The influence of these variables is not completely understood. The functional dependency between property changes and neutron radiation is summarized in the form of damage exposure parameters that are weighted integrals over the neutron fluence spectrum. The evaluation of neutron radiation effects on pressure vessel steels and the determination of safety limits require the knowlege of uncertainties in the prediction of radiation exposure parameters (for example, dpa (Practice E 693), neutron fluence greater than 1.0 MeV, neutron fluence greater than 0.1 MeV, thermal neutron fluence, etc.). This practice describes recommended procedures and data for determining these exposure parameters (and the associated uncertainties) for test reactor experiments. The nuclear industry draws much of its information from databases that come from test reactor experiments. Therefore, it is essential that reliable databases are obtained from test reactors to assess safety issues in Light Water Reactor (LWR) nuclear power plants.1.1 This practice covers the methodology summarized in Annex A1 to be used in the analysis and interpretation of physics-dosimetry results from test reactors. 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods. 1.3 Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. 1.4 This practice is directed towards the development and application of physics-dosimetry-metallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practices E 185, E 560, E 853, and E 1035, Guides E 900, E 2005, E 2006 and Test Method E 646. 1.5 This standard may involve hazardous materials, operations, and equipment. 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 Practice for Analysis and Interpretation of Physics Dosimetry Results for Test Reactors, E 706(II)

ASTM E944-08 反应堆监测时中子光谱调节法的应用的标准指南,E 706(IIA)

Adjustment methods provide a means for combining the results of neutron transport calculations with neutron dosimetry measurements (see Test Method E 1005 and NUREG/CR-5049) in order to obtain optimal estimates for neutron damage exposure parameters with assigned uncertainties. The inclusion of measurements reduces the uncertainties for these parameter values and provides a test for the consistency between measurements and calculations and between different measurements (see 3.3.3). This does not, however, imply that the standards for measurements and calculations of the input data can be lowered; the results of any adjustment procedure can be only as reliable as are the input data. Input Data and Definitions: The symbols introduced in this section will be used throughout the guide. Dosimetry measurements are given as a set of reaction rates (or equivalent) denoted by the following symbols: These data are, at present, obtained primarily from radiometric dosimeters, but other types of sensors may be included (see 4.1). The neutron spectrum (see Terminology E 170) at the dosimeter location, fluence or fluence rate Φ(E) as a function of neutron energy E, is obtained by appropriate neutronics calculations (neutron transport using the methods of discrete ordinates or Monte Carlo, see Guide E 482). The results of the calculation are customarily given in the form of k group fluences or fluence rates. where: Ej and Ej+1 are the lower and upper bounds for the j-th energy group, respectively. The reaction cross sections of the dosimetry sensors are obtained from an evaluated cross section file. The cross section for the i-th reaction as a function of energy E will be denoted by the following: Used in connection with the group fluences, Eq 2, are the calculated group-averaged cross sections σij. These values are defined through the following equation: Uncertainty information in the form of variances and covariances must be provided for all input data. Appropriate corrections must be made if the uncertainties are due to bias producing effects (for example, effects of photo reactions). Summary of the Procedures: An adjustment algorithm modifies the set of input data as defined in 3.2 in the following manner (adjusted quantities are indicated by a tilde, for example, ?/span>i): or for group fluence rates

Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E706 (IIA)

ANSI/ANS 2.27-2008 地震危害评估用核设备场所的调查标准

This standard provides criteria and guidelines for conducting geological, seismological, and geotechnical investigations needed to provide information to support seismic source characterization input to a probabilistic seismic hazard analysis (PSHA); evaluation of surface fault rupture hazard; site response analysis; and seismically-induced ground failure hazard. These criteria are applicable for SDC-3, SDC-4, and SDC-5 SSCs.

Criteria for Investigations of Nuclear Facility Sites for Seismic Hazard Assessments

ANSI/ANS 2.29-2008 地震危害可能性分析

This standard provides criteria to establish the probabilistic basis for various levels of natural phenomena hazards at nuclear materials facility sites.

Probabilistic Seismic Hazard Analysis

ASTM C1174-07 高放射性废弃物地质处置的工程屏障体系(EBS)用包括废弃物形状材料长期性能的预测用标准实施规程

1.1 This practice describes test methods and data analyses used to develop models for the prediction of the long-term behavior of materials, such as engineered barrier system (EBS) materials and waste forms, used in the geologic disposal of spent nuclear fuel (SNF) and other high-level nuclear waste in a geologic repository. The alteration behavior of waste form and EBS materials is important because it affects the retention of radionuclides by the disposal system. The waste form and EBS materials provide a barrier to release either directly (as in the case of waste forms in which the radionuclides are initially immobilized), or indirectly (as in the case of containment materials that restrict the ingress of groundwater or the egress of radionuclides that are released as the waste forms and EBS materials degrade).1.1.1 Steps involved in making such predictions include problem definition, testing, modeling, and model confirmation.1.1.2 The predictions are based on models derived from theoretical considerations, expert judgment, interpretation of data obtained from tests, and appropriate analogs. 1.1.3 For the purpose of this practice, tests 1.1.3 For the purpose of this practice, tests are categorized according to the information they provide and how it is used for model development and use. These tests may include but are not limited to the following:Attribute tests to measure intrinsic materials properties,Characterization tests to measure the effects of material and environmental variables on behavior,Accelerated tests to accelerate alteration and determine important mechanisms and processes that can affect the performance of waste form and EBS materials,Service condition tests to confirm the appropriateness of the model and variables for anticipated disposal conditions,Confirmation tests to verify the predictive capacity of the model, andTests or analyses performed with analog materials to identify important mechanisms, verify the appropriateness of an accelerated test method, and to confirm long-term model predictions.1.2 The purpose of this practice is to provide methods for developing models that can be used for the prediction of materials behavior over the long periods of time pertinent to the service life of a geologic repository as part of the basis for performance assessment of the repository.1.3 This practice also addresses uncertainties in materials behavior models and their impact on the confidence in the performance assessment.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 requirements prior to use.

Standard Practice for Prediction of the Long-Term Behavior of Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geological Disposal of High-Level Radioactive Waste

EJ/T 1201-2006 核燃料后处理厂退役设计安全准则

本标准规定了核燃料后处理厂退役设计的安全准则。 本标准适用于正常关闭的核燃料后处理厂的退役设计。非正常关闭的后处理厂以及放射性废物处理设施的退役设计也可参照使用。

Safety criteria for decommissioning design of spect fuel reprocessing plant

NF M60-810:2006 核素比环境测量用在线α-射线光谱测定法

In-situ gamma-ray spectrometry for nuclide specific environmental measurements.

NF M60-812-1:2006 核能.测量环境放射性.第1部分:用环境碳样品测量空气传播碳14测定体积活性指南

Nuclear Energy - Measurement of environmental radioactivity - Part 1 : guide for the measurement of airborne C14 volumetric activity from an atmospheric sample.

ANSI/ANS 6.4.2-2006 辐射屏蔽材料的规范

This standard sets forth physical and nuclear properties that shall be reported by the supplier as appropriate for a particular application in order to form the basis for the selection of radiation shielding materials

Radiation Shielding Materials, Specification for

NF M60-600:2006 辐射防护.监测处于放射性物质场维修风险环境中的职业人员

Radiation protection - Monitoring of workers occupationally exposed to a risk of internal contamination with radioactive material.

ISO 20553:2006 辐射防护.对职业性暴露到放射性物质内部污染风险环境中人员的监测

This International Standard specifies the minimum requirements for the design of professional programmes to monitor workers exposed to the risk of internal contamination by radioactive substances and establishes principles for the development of compatible goals and requirements for monitoring programmes. This International Standard addresses the a) purposes of monitoring and of monitoring programmes; b) description of the different categories of monitoring programmes; c) quantitative criteria for conducting monitoring programmes; d) suitable methods for monitoring and criteria for their selection; e) information that has to be collected for the design of a monitoring programme; f) general requirements for monitoring programmes (e.g. detection limits, tolerated uncertainties); g) frequencies of measurements; h) special cases; i) quality assurance; and j) documentation, reporting, record-keeping. This International Standard does not address - the monitoring of exposure to radon and its radioactive decay products; - detailed descriptions of measuring methods and techniques; - detailed procedures for in vivo measurements and in vitro analyses; - interpretation of monitoring results in terms of doses; - biokinetic data and mathematical models for converting measured activities into absorbed dose, equivalent dose and effective dose; or - the investigation of the causes or implications of an exposure or intake.

Radiation protection - Monitoring of workers occupationally exposed to a risk of internal contamination with radioactive material

EJ 1056-2005 铀加工与燃料制造设施辐射防护规定

本标准规定了铀加工及燃料制造(包括铀的精制与六氟化铀转化设施、铀浓缩设施和燃料制造设施)实践中应遵循的辐射防护、环境保护要求及有关剂量限值和措施等。 本标准适用于铀加工及燃料制造设施的选址、设计、建设、运行和退役。

Regulations for radiation protection for uranium processing and fuel fabrication facilities

EJ/T 623-2005 铀加工及燃料制造设施的职业照射监测规定

本标准规定了铀加工及燃料制造设施职业照射的监测原则、方法、实施及对监测结果评价的基本要求。 本标准适用于铀加工及燃料制造设施运行的职业照射监测。

Regulations for monitoring of occupational exposure for uranium processing and fuel fabrication facilities

EJ/T 1192-2005 生产堆退役源项调查取样技术准则

本标准规定了生产堆退役源项调查取样点设计和取样方法、样品预处理要求；还规定了取样过程中的辐射防护和质量保证等相关技术要求。 本标准适用于生产堆退役放射性源项调查取样活动。其他类型反应堆退役源项调查取样工作可参照使用。

Technical criteria for sampling of source term survey for decommissioning of production reactor