F40 核材料、核燃料综合 标准查询与下载

ASTM C1493-09 用微分衰减系统钝化-活化中子计数法对废物中的核物质进行无损化验的标准试验方法

This test method is useful for quantifying fissile (for example, 233U, 235U, 239Pu and 241Pu) and spontaneously-fissioning nuclei (for example, 238Pu, 240Pu, 242Pu, 244Cm, 248Cm, and 252Cf) in waste and scrap drums. Total elemental mass of the radioactive materials can be calculated if the relative abundances of each radionuclide are known. Typically, this test method is used to measure one fissile isotope (for example, 235U or 239Pu). This test method can be used to segregate low level and transuranic waste at the 100 nCi/g concentration level currently required to meet the DOE Waste Isolation Pilot Plant (WIPP) waste acceptance criterion (5, 8, 9). This test method can be used for waste characterization to demonstrate compliance with the radioactivity levels specified in waste, disposal, and environmental regulations (See NRC regulatory guides, DOE Order 435.1, 10 CFR Part 71, 40 CFR Part 191, and DOE /WIPP-069). In the active mode, the DDT system can measure the 235U content in the range from <0.02 to >100 g and the 239Pu content, nominally between <0.01 and >20 g. In the passive mode, the DDT system is capable of assaying spontaneously-fissioning nuclei, over a nominal range from 0.05 to 15 g of 240Pu, or equivalent (5, 10, 11, 12, 13). This test method should be used in conjunction with a waste management plan that segregates the contents of assay items into material categories according to some or all of the following criteria: bulk density of the waste, chemical forms of the plutonium or uranium and matrix, (α, n) neutron intensity, hydrogen (moderator) and absorber content, thickness of fissile mass(es), and the assay item container size and composition. Each matrix may require a different set of calibration standards and may have different mass calibration limits. The effect on the quality of the assay (that is, minimizing precision and bias) can significantly depend on the degree of adherence to this waste management plan. The bias of the measurement results is related to the fill height, the homogeneity and composition of the matrix, the quantity and distribution of the nuclear material, and the item size. The precision of the measurement results is related to the quantity of the nuclear material, the background, and the count time of the measurement. For both matrix-specific and wide-range calibrations, this test method assumes the calibration material matches the items to be measured with respect to homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity, distribution, and form of nuclear material, to the extent they affect the measurement. The algorithms for this test method assume homogeneity. Heterogeneity in the distribution of nuclear material, neutron moderators, and neutron absorbers has the potential to cause biased results (14). This test method assumes that the distribution of the contributing radioisotopes is uniform throughout the container and that lumps of nuclear material are not present. Reliable results from the application of this test method require waste to be packaged so the conditions of Section 5.5 can be met. In some cases, site-specific requirements will dictate the packaging requirements with possible detrimental effects to the measurement results. Both the active mode and the passive mode provide assay values for plutonium. During the calibration process, the operator should determine the applicable mass ranges for both modes of operation.1.1 This test method covers a sy......

Standard Test Method for Non-Destructive Assay of Nuclear Material in Waste by Passive and Active Neutron Counting Using a Differential Die-Away System

ANSI N15.8-2009 核材料控制方法用美国国家标准.材料控制设备.核电站用特殊核材料控制和核算设备

ANSI N15.8-1974 was written when the fuel assembly was the primary unit of interest for control and accounting of special nuclear material. At the time, there was no specific guidance on control and accounting of pieces - such as rod segments and loose fuel pellets - resulting from fuel damage. The revised standard will provide clearer guidelines for controlling and accounting for all special nuclear material. These guidelines will ensure the safe and secure use of special nuclear material and protect the health and safety of the public.

American National Standard for Methods of Nuclear Material Control; Material Control Systems - Special Nuclear Material Control and Accounting Systems for Nuclear Power Plants

ASTM C1682-09 核废料支援地学处置贮藏库的特性描述标准指南

In order to demonstrate conformance to regulatory requirements and support the postclosure repository performance assessment information is required about the attributes, characteristics, and behavior of the SNF need to be determined. These properties of the SNF in turn support the transport, interim storage, and repository preclosure safety analyses, and repository postclosure performance assessment. The interim dry storage of commercial LWR SNF is regulated per Code of Federal Regulations, Title 10, Part 72, which requires that the cladding must not sustain during the interim storage period any “gross” damage sufficient to release fuel from the cladding into the container environment. However, cladding damage insufficient to allow the release of fuel during the interim storage period may still occur in the form of small cracks or pinholes. These cracks/pinholes could be sufficient to classify the fuel as “ failed fuel” or “breached fuel” per the definitions given in Section 3 for repository disposal purposes, because they could allow contact of water vapor or liquid with the spent fuel matrix and thus provide a pathway for radionuclide release from the waste form. Also, pinholes/cracks in fuel rods in dry or wet interim storage can also develop into much larger defects (for example, the phenomenon of cladding “unzipping”) under long-term repository conditions. Therefore SNF characterization should be adequate to determine the amount of “failed fuel” for either usage as required. This could involve the examination of reactor operating records, ultrasonic testing, sipping, and analysis of the residual water and drying kinetics of the spent fuel assemblies or canisters. In the U.S. the disposal of spent nuclear fuel (SNF) and high level waste (HLW) in a geologic repository is regulated under the Code of Federal Regulations, Title 10, Part 60 (which pertains to any geologic repository built in the U.S.), Code of Federal Regulations, Title 10, Part 63 (which pertains to the repository located at Yucca Mountain, Nevada), and Code of Federal Regulations, Title 40, Part 191 (which established the radiation protection levels the repository must meet). Each of these regulations contains constraints and limitations on the chemical or physical (or both) properties and long-term degradation behavior of the spent fuel and HLW in the repository. Evaluating the design and performance of the WF, WP, and the rest of the engineered barrier system (EBS) with respect to these regulatory constraints requires knowledge of the chemical/physical characteristics and degradation behavior of the SNF that could be provided by the testing and data evaluation methods provided by this guide, as follows: Code of Federal Regulations, Title 10, Part 60 Sections 135 and 113 require that the waste form be a material that is solid, non-particulate, non-pyrophoric, and non-chemically reactive, that the waste package contain no liquid, particulates, or combustible materials and that the materials/components of the EBS be designed to provide—assuming anticipated processes and events—substantially complete containment of the HLW for the NRC-designated regulatory period. Code of Federal Regulations, Title 10, Part 63 Section 113 requires that the EBS be designed such that, working in combination with the natural barriers, the performance assessment of the EBS demonstrates conformance to the annual reasonabl......

Standard Guide for Characterization of Spent Nuclear Fuel in Support of Geologic Repository Disposal

ISO 9278:2008 核能.丸状二氧化铀.开口孔隙率和封闭孔隙率密度和体积分率的测定

Nuclear energy - Uranium dioxide pellets - Determination of density and volume fraction of open and closed porosity

ISO 21484:2008 核燃料技术.MOX芯块中O/M比率的测定.重量分析法

Nuclear fuel technology - Determination of the O/M ratio in MOX pellets - Gravimetric method

GJB 6631-2008 军用核材料衡算中体积校准技术导则

本标准规定了军用核材料衡算中计量槽体积校准和在槽液体体积测量的方法。 本标准适用于军用材料衡算中配备压差测量系统的计量槽。

Guide for volume calibration techniques of military nuclear materials accounting

GJB 6632-2008 军用核材料衡算中质量测量校准技术导则

本标准规定了军用核材料衡算中表征质量测量过程、测定试验物体的质量值及建立测量保证方案等内容。 本标准适用于： a)选择衡器和质量标准。 b)评价质量测量过程的性能，包括估计计随机误差、系统误差和误差限值的大小； c)测定试验物体的质量值； d)使用仿制质量标准（RMS），建立和维护一个对设施内和设施间质量测量进行控制的方案。

Guide for mass calibration techniques of military nuclear materials accounting

GJB 6633-2008 离心级联核材料实物盘存技术要求

本标准规定了离心工厂实物盘存的方式与周期、组织、准备、实施、检验等内容。 本标准适用于铀浓缩离心工厂的离心级联、供取料厂房、样品分析实验室、液化倒料、废料回收等核材料实物盘存。 本标准不包括库房盘存。

Requirement for centrifuge cascade nuclear material physical inventory taking

EJ/T 1230-2008 地浸砂岩型铀矿镭氡平衡系数测量规程

本标准规定了地浸砂岩型铀矿镭氡平衡破坏的检查、镭氡平衡系数测量方法和技术要求、修正方法以及质量要求等内容。 本标准适用于地浸砂岩型铀矿勘查各阶段镭氡平衡系数的测定，也可作为野外原始资料验收和各勘查阶段地质报告、资源/储量报告评审的依据.

Measurement rules on radium and radon equilibrium coefficient for in-situ leaching sandstone type uranium deposits

ISO 18213-4:2008 核燃料技术.核材料衡算用容器校准和体积测定.第4部分:装备汲取管的衡算容器中液体高度的精确测定、低速起泡率

This part of ISO 18213 specifies a procedure for making accurate determinations of the liquid height in nuclear-materials-accountancy tanks that are equipped with pneumatic systems for determining the liquid content. With such systems, gas is forced through a probe (dip tube) whose tip is submerged in the tank liquid. The pressure required to induce bubbling is measured with a manometer located at some distance from the tip of the probe. This procedure applies specifically when a very slow bubbling rate is employed. A series of liquid height determinations made with a liquid of known density is required to estimate a tank's calibration equation (see ISO 18213-1), the function that relates the elevation (height) of a point in the tank to an independent determination of tank volume associated with that point. For accountability purposes, the tank's measurement equation (the inverse of its calibration equation) is used to determine the volume of process liquid in the tank that corresponds to a given determination of the liquid height.

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 4: Accurate determination of liquid height in accountancy tanks equipped with dip tubes, slow bubbling rate

ISO 18213-5:2008 核燃料技术.核材料衡算用容器校准和体积测定.第5部分:装备汲取管的衡算容器中液体高度的精确测定、快速起泡率

This part of ISO 18213 specifies a procedure for making accurate determinations of liquid height in nuclear- materials-accountancy tanks that are equipped with pneumatic systems for determining the liquid content. With such systems, gas is forced through a probe (dip tube) whose tip is submerged in the tank liquid. The pressure required to induce bubbling is measured with a manometer located at some distance from the tip of the probe. This procedure applies specifically when a fast bubbling rate is employed. A series of liquid height determinations made with a liquid of known density is required to estimate a tank's calibration equation (see ISO 18213-1), the function that relates the elevation (height) of a point in the tank to an independent determination of tank volume associated with that point. For accountability purposes, the tank's measurement equation (the inverse of its calibration equation) is used to determine the volume of process liquid in the tank that corresponds to a given determination of liquid height.

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 5: Accurate determination of liquid height in accountancy tanks equipped with dip tubes, fast bubbling rate

ISO 18213-6:2008 核燃料技术 核材料衡算用容器校准和体积测定 第6部分 装备有汲取管的衡算容器中液体精确密度的容器内测定

This part of ISO 18213 specifies a procedure for making accurate determinations of the densities of process liquids from in-tank measurements of the liquid content. This procedure is applicable to tanks equipped with pneumatic systems for determining the liquid content that have two or more bubbler probes of differing lengths. It is necessary that the probes be fixed relative to each other and to the tank in which they are installed. The methods presented in this part of ISO 18213 yield acceptable results only for clear (i.e. free of suspended solids) liquids that are both homogeneous in concentration and at thermal equilibrium. The accuracy of the method is limited by -- the accuracy of the density determinations for the calibration liquid, and -- the number and accuracy of the liquid height determinations used in the calculations. With state-of-the-art measurement technology and careful technique, and with water as a calibration liquid, it is possible to determine the density of homogeneous clear liquids with relative accuracy1) on the order of 2 x 10 to 3 x 10, or approximately 0,2 kg/m3 or 0,3 kg/m3.

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 6: Accurate in-tank determination of liquid density in accountancy tanks equipped with dip tubes

DIN 25463-2:2008 轻水反应堆的核燃料衰变功率的计算.第2部分:压水反应堆用铀、钚混合氧化物(MOX)

Calculation of the decay power in nuclear fuels of light water reactors - Part 2: Mixed-uranium-plutonium oxide (MOX) nuclear fuel for pressurized water reactors

NF M60-430-2*NF ISO 18213-2:2008 核燃料技术.核原料衡算用容器的校准和体积的测定.第2部分:容器校准数据的标准化

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 2 : data standardization for tank calibration.

ISO 22875:2008 核能.二氧化铀粉末和烧结颗粒中氯和氟的测定

This International Standard describes a method for determining chlorine and fluorine in uranium dioxide powder and sintered pellets. It is applicable for the analysis of samples with a mass fraction of chlorine from 5 ug/g to 500 ug/g and with a mass fraction of fluorine from 2 ug/g to 500 ug/g.

Nuclear energy - Determination of chlorine and fluorine in uranium dioxide powder and sintered pellets

ASTM C1346-08 P-10管中UF 6溶解的标准实施规程

Uranium hexafluoride is a basic material used to prepare nuclear reactor fuel. To be suitable for this purpose the material must meet criteria for uranium content, isotopic composition and metallic impurities in Specification C 787 and C 996. This practice results in the complete dissolution of the sample for uranium and impurities analysis, and determination of isotopic distribution by mass spectrometry as described in, for example, Test Methods C 761.1.1 This practice covers the dissolution of UF6 from a P-10 tube to provide solutions for analysis. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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. For specific safeguard and safety precaution statements, see Section 8.

Standard Practice for Dissolution of UF₆ from P-10 Tubes ^,

NF M60-462-1*NF ISO 21847-1:2007 核燃料技术.α光谱测定法.第1部分:铀及其化合物中镎的测定

Nuclear fuel technology - Alpha spectrometry - Part 1 : determination of neptunium in uranium and its compounds.

NF M60-462-3*NF ISO 21847-3:2007 核燃料技术.α光谱测定法.第3部分:铀及其化合物中铀232的测定

Nuclear fuel technology - Alpha spectrometry - Part 3 : determination of uranium 232 in uranium and its compounds.

ISO 18213-2:2007 核燃料技术.核材料衡算用容器校准和体积测定.第2部分:容器校准数据的标准化

This part of ISO 18213 presents procedures for standardizing a set of calibration data to a fixed set of reference conditions so as to minimize the effect of variations in ambient conditions that occur during the measurement process. The procedures presented herein apply generally to measurements of liquid height and volume obtained for the purpose of calibrating a tank (i.e. calibrating a tank’s measurement system). When used in connection with other parts of ISO 18213, these procedures apply specifically to tanks equipped with bubbler probe systems for measuring liquid content. The standardization algorithms presented herein can be profitably applied when only estimates of ambient conditions, such as temperature, are available. However, the most reliable results are obtained when relevant ambient conditions are measured for each measurement of volume and liquid height in a set of calibration data.

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 2: Data standardization for tank calibration

ISO 18213-1:2007 核燃料技术.核材料衡算用容器校准和体积测定.第1部分:程序综述

This part of ISO 18213 describes procedures for tank calibration and volume determination for nuclear process tanks equipped with pressure-measurement systems for determining liquid content. Specifically, overall guidance is provided for planning a calibration exercise undertaken to obtain the data required for the measurement equation to estimate a tank’s volume. The key steps in the procedure are also presented for subsequently using the estimated volume-measurement equation to determine tank liquid volumes. The procedures presented apply specifically to tanks equipped with bubbler probe systems for measuring liquid content. Moreover, these procedures produce reliable results only for clear (i.e. without suspended solids), homogeneous liquids that are at both thermal and static equilibrium.

Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 1: Procedural overview