F46 核材料、核燃料及其分析试验方法 标准查询与下载

ASTM C1267-11 使用铁(II)在磷酸中还原后在有钒的情况下滴定铬(VI)的铀标准试验方法

Factors governing selection of a method for the determination of uranium include available quantity of sample, homogeneity of material sampled, sample purity, desired level of reliability, and facility available equipment. This uranium assay method is referenced in the Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets (Test Methods C696) and in the Test Methods for Chemical, Mass Spectrometric, and Spectrochemical, Nuclear, and Radiochemical Analysis of Nuclear-Grade Uranyl Nitrate Solutions (Test Methods C799). This uranium assay method may also be used for uranium hexafluoride and uranium ore concentrate. This test method determines 20 to 200 mg of uranium; is applicable to product, fuel, and scrap material after the material is dissolved; is tolerant towards most metallic impurity elements usually specified in product and fuel; and uses no special equipment. The ruggedness of the titration method has been studied for both the volumetric (6) and the weight (7) titration of uranium with dichromate. Committee C26 Safeguards Statement: The materials (nuclear grade uranium in product, fuel, and scrap) to which this test method applies are subject to nuclear safeguard regulations governing their possession and use. The analytical method in this standard meets U.S. Department of Energy guidelines for acceptability of a measurement method for generation of safeguards accountability measurement data. When used in conjunction with the appropriate certified reference materials (SRM or CRM), this procedure can demonstrate traceability to the national measurement base. However, use of the test method does not automatically guarantee regulatory acceptance of the resulting safeguards measurements. It remains the sole responsibility of the user of this test method to assure that its application to safeguards has the approval of the proper regulatory authorities.1.1 This test method, commonly referred to as the Modified Davies and Gray technique, covers the titration of uranium in product, fuel, and scrap materials after the material is dissolved. The test method is versatile and has been ruggedness tested. With appropriate sample preparation, this test method can give precise and unbiased uranium assays over a wide variety of material types (1, 2). Details of the titration procedure in the presence of plutonium with appropriate modifications are given in Test Method C1204. 1.2 Uranium levels titrated are usually 20 to 50 mg, but up to 200 mg uranium can be titrated using the reagent volumes stated in this test method. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific safeguard and safety precaution statements, see Section 4.

Standard Test Method for Uranium by Iron (II) Reduction in Phosphoric Acid Followed by Chromium (VI) Titration in the Presence of Vanadium

ASTM D7301-11 适用于受低中子辐照剂量元件的核石墨标准规范

1.1 This specification covers the classification, processing, and properties of nuclear grade graphite billets with dimensions sufficient to meet the designer’s requirements for reflector blocks and core support structures, in a high temperature gas cooled reactor. The graphite classes specified here would be suitable for reactor core applications where neutron irradiation induced dimensional changes are not a significant design consideration. 1.2 The purpose of this specification is to document the minimum acceptable properties and levels of quality assurance and traceability for nuclear grade graphite suitable for components subjected to low irradiation dose. Nuclear graphites meeting the requirements of Specification D7219 are also suitable for components subjected to low neutron irradiation dose. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

Standard Specification for Nuclear Graphite Suitable for Components Subjected to Low Neutron Irradiation Dose

ASTM C696-11 对核级二氧化铀粉末和球粒进行化学分析,质谱分析及光谱分析的标准试验方法

Uranium dioxide is used as a nuclear-reactor fuel. In order to be suitable for this purpose, the material must meet certain criteria for uranium content, stoichiometry, isotopic composition, and impurity content. These test methods are designed to show whether or not a given material meets the specifications for these items as described in Specifications C753 and C776. An assay is performed to determine whether the material has the minimum uranium content specified on a dry weight basis. The stoichiometry of the oxide is useful for predicting its sintering behavior in the pellet production process. Determination of the isotopic content of the uranium in the uranium dioxide powder is made to establish whether the effective fissile content is in compliance with the purchaser''s specifications. Impurity content is determined to ensure that the maximum concentration limit of certain impurity elements is not exceeded. Determination of impurities is also required for calculation of the equivalent boron content (EBC).1.1 These test methods cover procedures for the chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade uranium dioxide powders and pellets to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Uranium by Ferrous Sulfate Reduction in Phosphoric Acid and Dichromate Titration Method C1267 Test Method for Uranium By Iron (II) Reduction In Phosphoric Acid Followed By Chromium (VI) Titration In The Presence of Vanadium Uranium and Oxygen Uranium Atomic Ratio by the Ignition (Gravimetric) Impurity Correction Method C1453 Standard Test Method for the Determination of Uranium by Ignition and Oxygen to Uranium Ratio (O/U) Atomic Ratio of Nuclear Grade Uranium Dioxide Powders and Pellets Carbon (Total) by Direct Combustion-Thermal Conductivity Method C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Total Chlorine and Fluorine by Pyrohydrolysis Ion-Selective Electrode Method C1502 Standard Test Method for the Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide nbsp;nbsp;nbsp; Moisture by the Coulometric, Electrolytic Moisture Analyzer Method7-14 Nitrogen by the Kjeldahl Method15-22

Standard Test Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets

ANSI/ANS 5.4-2011 氧化物燃料中挥发裂变产物的释放比例计算方法

This standard provides an analytical method for calculating the release of volatile fission products from uranium dioxide fuel pellets during normal reactor operation. When used with nuclide yields, this method will give the release-to-birth ratio, R/B, or the so-called gap release, which is the inventory of volatile radioactive fission products that could be available for release from the fuel rod if the cladding were breached.

Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel

ASTM C889-11 核纯级的氧化钆(Gd2O3)粉末的化学和质谱分析标准试验方法

Gadolinium oxide powder is used, with subsequent processing, in nuclear fuel applications, such as an addition to uranium dioxide. These test methods are designed to determine whether the material meets the requirements described in Specification C888. The material is analyzed to determine whether it contains the minimum gadolinium oxide content specified. The loss on ignition and impurity content are determined to ensure that the weight loss and the maximum concentration limit of specified impurity elements are not exceeded.1.1 These test methods cover procedures for the chemical and mass spectrometric analysis of nuclear-grade gadolinium oxide powders to determine compliance with specifications. 1.2 The analytical procedures appear in the following order: Sections Carbon by Direct CombustionThermal Conductivity C1408 Test Method for Carbon (Total) in Uranium Oxide Powders and Pellets By Direct Combustion-Infrared Detection Method Total Chlorine and Fluorine by Pyrohydrolysis Ion Selective Electrode C1502 Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide Loss of Weight on Ignition7-13 Sulfur by CombustionIodometric Titration Impurity Elements by a Spark-Source Mass Spectrographic C761 Test Methods for Chemical, Mass Spectrometric, Spectrochemical,Nuclear, and Radiochemical Analysis of Uranium Hexafluoride C1287 Test Method for Determination of Impurities In Uranium Dioxide By Inductively Coupled Plasma Mass Spectrometry Gadolinium Content in Gadolinium Oxide by Impurity Correction14-17 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazard statements, see Section 5. 7.1 This test method covers the loss-on-ignition of volatile constituents from nuclear-grade gadolinium oxide (Gd2O3) powder. 14.1 The percent gadolinium oxide content of powders, exclusive of volatiles, is determined by calculation after the material......

Standard Test Methods for Chemical and Mass Spectrometric Analysis of Nuclear-Grade Gadolinium Oxide (Gd₂O₃) Powder

ASTM C1031-11(2016) 核级氧化铝粉的标准规格

1.1 This specification provides the chemical and physical requirements for nuclear-grade aluminum oxide powder intended for fabrication into shapes for nuclear applications. Two specific uses for which this powder is intended are Al2O3 pellets and Al2O38201;−8201;B4C composite pellets for use as thermal insulator or burnable neutron absorbers, respectively. 1.2 The material described herein shall be particulate in nature.

Standard Specification for Nuclear-Grade Aluminum Oxide Powder

ASTM C761-11 六氟化铀的化学,质谱,光谱化学,核(放射性)及放射化学分析的标准试验方法

1.1 These test methods cover procedures for subsampling and for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of uranium hexafluoride UF6. Most of these test methods are in routine use to determine conformance to UF6 specifications in the Enrichment and Conversion Facilities.1.2 The analytical procedures in this document appear in the following order:Note 18212;Subcommittee C26.05 will confer with C26.02 concerning the renumbered section in Test Methods C761 to determine how concerns with renumbering these sections, as analytical methods are replaced with stand-alone analytical methods, are best addressed in subsequent publications.SectionsSubsampling of Uranium Hexafluoride7 - 10Gravimetric Determination of Uranium11 - 19Titrimetric Determination of Uranium20 Preparation of High-Purity U3O 821Isotopic Analysis22Isotopic Analysis by Double-Standard Mass-Spectrometer Method23 - 29Determination of Hydrocarbons, Chlorocarbons, and Partially Substituted Halohydrocarbons29-36Atomic Absorption Determination of Antimony36Spectrophotometric Determination of Bromine37Titrimetric Determination of Chlorine38-44Determination of Silicon and Phosphorus45-51Determination of Boron and Silicon52-59 Determination of Ruthenium60 Determination of Titanium and Vanadium61Spectrographic Determination of Metallic Impurities 62Determination of Tungsten63Determination of Thorium and Rare Earths64-69Determination of Molybdenum70Atomic Absorption Determination of Metallic Impurities71-76Impurity Determination by Spark-Source Mass Spectrography77Determination of Boron-Equivalent Neutron Cross Section78Determination of Uranium-233 Abundance by Thermal Ionization Mass Spectrometry79Determination of Uranium-232 by Alpha Spectrometry80-86Determination of Fission Product Activity87Determination of Plutonium by Ion Exchange and Alpha Counting88-92Determination of Plutonium by Extraction and Alpha Counting93-100Determination of Neptunium by Extraction and Alpha Counting101-108Atomic Absorption Determination of Chromium Soluble In Uranium Hexafluoride109-115Atomic Absorption Determination of Chromium Insoluble In Uranium Hexafluoride116-122Determination of Technetium-99 In Uranium Hexafluoride123-131Method for the Determiation of Gamma-Energy Emission Rate from Fission Products in Uranium Hexafluoride132Determination of Metallic Impurities by ICP-AES133-142Determination of Molybdenum, Niobium, Tantalum, Titanium, and Tungsten by ICP-AES143-1521.3 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 consideration statements, see Section 6.)

Standard Test Methods for Chemical, Mass Spectrometric, Spectrochemical, Nuclear, and Radiochemical Analysis of Uranium Hexafluoride

ASTM C1234-11 痕量元素测定用高压,高温分解法制备油及含油废物样品的标准操作规程

This practice is useful for preparation of difficult-to-digest, primarily oils and oily wastes, specimens for trace element determinations of up to 28 elements by atomic absorption or plasma emission techniques. Specimen preparation by high-pressure ashing is primarily applicable to specimens whose preparation by EPA SW-846 protocols is either not applicable or not defined. This sample preparation practice is applicable for the trace element characterization of mixed oily wastes for use by waste treatment facilities such as incinerators or waste stabilization facilities.1.1 This practice covers a high-pressure, high-temperature digestion technique using the high-pressure asher (HPA) for preparation of oils and oily waste specimens for determination of up to 28 different elements by inductively coupled plasma-atomic emission plasma spectroscopy (ICP-AES), cold-vapor atomic absorption spectroscopy (CVAAS), and graphite furnace atomic absorption spectroscopy (GFAAS), inductively coupled plasma-mass spectrometry (ICPMS), and radiochemical methods. Oily and high-percentage organic waste streams from nuclear and non-nuclear manufacturing processes can be successfully prepared for trace element determinations by ICP-AES, CVAAS, and GFAAS. This practice is applicable to the determination of total trace elements in these mixed wastes. Specimens prepared by this practice can be used to characterize organic mixed waste streams received by hazardous waste treatment incinerators and for total element characterization of the waste streams. 1.2 This practice is applicable only to organic waste streams that contain radioactivity levels that do not require special personnel or environmental protection from radioactivity or other acute hazards. 1.3 A list of elements determined in oily waste streams is found in Table 1. 1.4 This practice has been used successfully to completely digest a large variety of oils and oily mixed waste streams from nuclear processing facilities. While the practice has been used to report data on up to 28 trace elements, its success should not be expected for all analytes in every specimen. The overall nature of these oily wastes tends to be heterogeneous that can affect the results. Homogeneity of the prepared sample is critical to the precision and quality of the results. 1.5 This practice is designed to be applicable to samples whose preparation practices are not defined, or not suitable, by other regulatory procedures or requirements, such as the U.S. Environmental Protection Agency (EPA) SW-846 and EPA-600/4-79-020 documents. This digestion practice is designed to provide a high level of accuracy and precision, but does not replace or override any regulatory requirements for sample preparation. 1.6 This practice uses hazardous materials, operations, and equipment at high pressure (90–110 bars, 89–108 atm, or 1305–1595 lb/in.2) and high temperatures, up to 320°C, and therefore poses significant hazards if not operated properly. 1.7 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.7.1 Exception8212;Pressure measurements are given in lb/in. units. 1.8 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. Specific warning statements are given in Sections 10. TABLE 1 List of......

Standard Practice for Preparation of Oils and Oily Waste Samples by High-Pressure, High-Temperature Digestion for Trace Element Determinations

NF M60-469*NF ISO 26062:2010 核工艺学.核燃料.用电感耦合等离子体质谱法测量铀和钚基材料中元素状态杂质的程序.

Nuclear technology - Nuclear fuels - Procedures for the measurement of elemental impurities in uranium- and plutonium based materials by inductively coupled plasma mass spectrometry.

BS ISO 26062:2010 核工艺学.核燃料.用电感耦合等离子体质谱法测量铀和钚基材料中元素状态杂质的规程.

Nuclear technology. Nuclear fuels. Procedures for the measurement of elemental impurities in uranium- and plutonium-based materials by inductively coupled plasma mass spectrometry

ISO 26062:2010 核技术.核燃料.使用电感耦合等离子体质谱法测量铀和钚基础材料中杂质元素的规程

This International Standard specifies a procedure for the determination of trace impurities in uranium- or plutonium-based, or mixed uranium- and plutonium-based, materials by inductively coupled plasma mass spectrometry (ICP-MS). It provides both guidelines and specific options for the determination of an element or group of elements. It is applicable to solutions such as uranyl or plutonium nitrate, solids such as the oxides and to mixed actinide materials such as unirradiated mixed oxide material in either solid or dissolved forms. It is not directly suitable for the analysis of uranium or plutonium matrices containing significant quantities of other elements such as uranium–gadolinium mixtures. It may nevertheless form the basis of a process for analysing this type of matrix, provided that the impact of the gadolinium component is ascertained.

Nuclear technology - Nuclear fuels - Procedures for the measurement of elemental impurities in uranium- and plutonium-based materials by inductively coupled plasma mass spectrometry

BS ISO 9005:2007 核能.二氧化铀粉末和烧结芯块.用安培法测定氧/铀原子比

Nuclear energy - Uranium dioxide powder and sintered pellets - Determination of oxygen/uranium atomic ratio by the amperometric method

ISO 10276:2010 核能.燃料技术.用于运送放射性材料的包装用耳筒

This International Standard covers trunnions fitted to radioactive-material transport packages that are subject to the approval and licensing by competent authorities in accordance with the IAEA No. TS-R-1. Aspects included are design, manufacture, maintenance and quality assurance. Subject to agreement between the interested parties, it can also be applied to packages that are not subject to the approval by competent authorities. This International Standard covers trunnion systems used for tie-down during transport and trunnions used for tilting and/or lifting. This International Standard does not supersede any of the requirements in the IAEA No. TS-R-1, nor any of the requirements of international or national regulations, concerning trunnions used for lifting and tie-down. This International Standard is applicable to new package design.

Nuclear energy - Fuel technology - Trunnions for packages used to transport radioactive material

ASTM C1742-10 用双标准单收集器气体质谱仪法进行六氟化铀同位素分析的标准试验方法

Uranium hexafluoride is a basic material used to prepare nuclear reactor fuel. To be suitable for this purpose, the material shall meet the criteria for isotopic composition. This test method is designed to determine whether the material meets the requirements described in Specifications C787 and C996. ASTM Committee C26 Safeguards Statement: The material (uranium hexafluoride) to which this test method applies is subject to the nuclear safeguards regulations governing its possession and use. The analytical procedure in this test method has been designated as technically acceptable for generating safeguards accountability data. When used in conjunction with the appropriate certified reference materials (CRMs), this procedure can demonstrate traceability to the national measurement base. However, adherence to this procedure does not automatically guarantee regulatory acceptance of the regulatory safeguards measurements. It remains the sole responsibility of the user of this test method to ensure that its application to safeguards has the approval of the proper regulatory authorities.1.1 This is a quantitative test method applicable to determining the mass percent of uranium isotopes in uranium hexafluoride (UF6) samples with 235U concentrations between 0.1 and 5.0 mass %. 1.2 This test method may be applicable for the entire range of 235U concentrations for which adequate standards are available. 1.3 This test method is for analysis by a gas magnetic sector mass spectrometer with a single collector using interpolation to determine the isotopic concentration of an unknown sample between two characterized UF6 standards. 1.4 This test method is to replace the existing test method currently published in Test Methods C761 and is used in the nuclear fuel cycle for UF6 isotopic analyses. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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 Isotopic Analysis of Uranium Hexafluoride by Double Standard Single-Collector Gas Mass Spectrometer Method

ASTM C1562-10 临时废核燃料干燥储存系统长期运行过程中使用的材料的评估的标准指南

Information is provided in this document and other referenced documents to assist the licensee and the licensor in analyzing the materials aspects of performance of SNF and DCSS components during extended storage. The effects of the service conditions of the first licensing period are reviewed in the license renewal process. These service conditions are highlighted and discussed in Annex A1 as factors that affect materials performance in an ISFSI. Emphasis is on the effects of time, temperature, radiation, and the environment on the condition of the SNF and the performance of components of ISFSI storage systems. The storage of SNF that is irradiated under the regulations of 10 CFR Part 50 is governed by regulations in 10 CFR Part 72. Regulatory requirements for the subsequent geologic disposal of this SNF are presently given in 10 CFR Part 60, with specific requirements for the use of Yucca Mountain as a repository being given in the regulatory requirements of 10 CFR Part 63. Between the life-cycle phases of storage and disposal, SNF may be transported under the requirements of 10 CFR Part 71. Therefore, in storage, it is important to acknowledge the transport and disposal phases of the life cycle. In doing this, the materials properties that are important to these subsequent phases are to be considered in order to promote successful completion of these subsequent phases in the life cycle of SNF. Retrievability of SNF (or high-level radioactive waste) is set as a requirement in 10 CFR Part 72.122(g)(5) and 10 CFR Part 72.122(l). Care should be taken in operations conducted prior to disposal, for example, storage, transfer, and transport, to ensure that the SNF is not abused and that SNF assemblies will be retrievable, the protective value of the cladding is not degraded and remains capable of serving as an active barrier to radionuclide release during transfer and transport operations. It is possible that cladding could be altered during dry storage. The hydrogen effects, fracture toughness of the cladding and the creep behavior are important parameters to be evaluated and controlled during the dry storage phase of the life cycle. These degradation mechanisms are discussed in Annex A2 and Annex A4.1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of this guide is based, in part, on the requirements for a dry SNF storage license that is granted, by the U.S. Nuclear Regulatory Commission (NRC), for up to 20 years. Although government regulations may differ for various nations, the guidance on materials properties and behavior given here is expected to have broad applicability. 1.2 This guide addresses many of the factors affecting the time-dependent behavior of materials under ISFSI service [10 CFR Part 72.42]. These factors are those regarded to be important to performance, in license extension, beyond the currently licensed 20-year period. Examples of these factors are given in this guide and they include materials al......

Standard Guide for Evaluation of Materials Used in Extended Service of Interim Spent Nuclear Fuel Dry Storage Systems

ASTM C859-10 与核材料相关用标准术语

1.1 This terminology standard contains terms, definitions, descriptions of terms, nomenclature, and explanations of acronyms and symbols specifically associated with standards under the jurisdiction of Committee C26 on Nuclear Fuel Cycle. This terminology may also be applicable to documents not under the jurisdiction of Committee C26, in which case this terminology may be referenced in those documents.

Standard Terminology Relating to Nuclear Materials

ASTM C1221-10 γ射线光谱法无损分析均相溶液中特种核材料的标准试验方法

This test method is a nondestructive means of determining the nuclide concentration of a solution for special nuclear material accountancy, nuclear safety, and process control. It is assumed that the nuclide to be analyzed is in a homogeneous solution (Practice C1168). The transmission correction makes the test method independent of matrix (solution elemental composition and density) and useful over several orders of magnitude of nuclide concentrations. However, a typical configuration will normally span only two to three orders of magnitude because of detector dynamic range. The test method assumes that the solution-detector geometry is the same for all measured items. This can be accomplished by requiring that the liquid height in the sidelooking geometry exceeds the detector field of view defined by the collimator. For the upward-looking geometry, a fixed solution fill height must be maintained and vials of identical radii must be used unless the vial radius exceeds the field of view defined by the collimator. Since gamma-ray systems can be automated, the test method can be rapid, reliable, and not labor intensive. This test method may be applicable to in-line or off-line situations.1.1 This test method covers the determination of the concentration of gamma-ray emitting special nuclear materials dissolved in homogeneous solutions. The test method corrects for gamma-ray attenuation by the solution and its container by measurement of the transmission of a beam of gamma rays from an external source (Refs. (1), (2), and (3)). 1.2 Two solution geometries, slab and cylinder, are considered. The solution container that determines the geometry may be either a removable or a fixed geometry container. This test method is limited to solution containers having walls or a top and bottom of equal transmission through which the gamma rays from the external transmission correction source must pass. 1.3 This test method is typically applied to radionuclide concentrations ranging from a few milligrams per litre to several hundred grams per litre. The assay range will be a function of the specific activity of the nuclide of interest, the physical characteristics of the solution container, counting equipment considerations, assay gamma-ray energies, solution matrix, gamma-ray branching ratios, and interferences. 1.4 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 appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific hazards, see Section 9.

Standard Test Method for Nondestructive Analysis of Special Nuclear Materials in Homogeneous Solutions by Gamma-Ray Spectrometry

ASTM C1379-10 电感耦合等离子体-质谱分析法对尿进行铀-235和铀-238分析的标准试验方法

DOE Order 5480.11 and ANSI N13.30 require that internal dose assessments be made as part of the bioassay program for nuclear facility workers. For indirect bioassay of uranium workers, the uranium isotopes must be measured along with the total uranium in urine samples. The RMDA for each uranium isotope is 0.1 pCi/L. This method is applicable for measuring 235U and 238U at the RMDA. Because of extremely low mass concentration (because of the high specific activity), 234U cannot be measured without additional sample preconcentration. Note 28212;Column chromatography separations and concentration of 234U using manual or flow-injection preconcentration followed by ICP-MS isotopic determination are described in Test Methods C1310 and C1345. These methods focus on environmental soil sample analysis, but with some development, may be applicable to digested urine samples. The 234U concentration can be calculated based on an enrichment gradient for workers in uranium enrichment plants, and internal dose assessments can be made. Note 38212;Use of high resolution ICP-MS may also be used to obtain lower detection limits. 1.1 This test method covers the determination of the concentration of uranium-235 and uranium-238 in urine using Inductively Coupled Plasma-Mass Spectrometry. This test method can be used to support uranium facility bioassay programs. 1.2 This method detection limits for 235U and 238U are 6 ng/L. To meet the requirements of ANSI N13.30, the minimum detectable activity (MDA) of each radionuclide measured must be at least 0.1 pCi/L (0.0037 Bq/L). The MDA translates to 47 ng/L for 235U and 300 ng/L for 238U. Uranium– 234 cannot be determined at the MDA with this test method because of its low mass concentration level equivalent to 0.1 pCi/L. 1.3 The digestion and anion separation of urine may not be necessary when uranium concentrations of more than 100 ng/L are present. 1.4 Units8212;The values stated in picoCurie per liter units are to be regarded as standard. The values given in parentheses are mathematical conversions to SI units that are provided for information only and are not considered 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. Note 18212;Warning: The ICP-MS is a source of intense ultraviolet radiation from the radio frequency induced plasma. Protection from radio frequency radiation and UV radiation is provided by the instrument under normal operation.

Standard Test Method for Analysis of Urine for Uranium-235 and Uranium-238 Isotopes by Inductively Coupled Plasma-Mass Spectrometry

ASTM C1718-10 使用层析成象γ扫描法对放射性物质进行无损分析的标准试验方法

The TGS provides a nondestructive means of mapping the attenuation characteristics and the distribution of the radionuclide content of items on a voxel by voxel basis. Typically in a TGS analysis a vertical layer (or segment) of an item will be divided into a number of voxels. By comparison, a segmented gamma scanner (SGS) can determine matrix attenuation and radionuclide concentrations only on a segment by segment basis. It has been successfully used to quantify 238Pu, 239Pu, and 235U. SNM loadings from 0.5 g to 200 g of 239Pu (5, 6), from 1 g to 25 g of 235U (7), and from 0.1 to 1 g of 238Pu have been successfully measured. The TGS technique has also been applied to assaying radioactive waste generated by nuclear power plants (NPP). Radioactive waste from NPP is dominated by activation products (for example, 54Mn, 58Co, 60Co, 110mAg) and fission products (for example, 137Cs, 134Cs). The radionuclide activities measured in NPP waste is in the range from 3.7E+04 Bq to 1.0E+07 Bq. Some results of TGS application to non-SNM radionuclides can be found in the literature (8). The TGS technique is well suited for assaying items that have heterogeneous matrices and that contain a non-uniform radionuclide distribution. Since the analysis results are obtained on a voxel by voxel basis, the TGS technique can in many situations yield more accurate results when compared to other gamma ray techniques such as SGS. In determining the radionuclide distribution inside an item, the TGS analysis explicitly takes into account the cross talk between various vertical layers of the item. The TGS analysis technique uses a material basis set method that does not require the user to select a mass attenuation curve apriori, provided the transmission source has at least 2 gamma lines that span the energy range of interest. A commercially available TGS system consists of building blocks that can easily be configured to operate the system in the SGS mode or in a far-field geometry. The TGS provides 3-dimensional maps of gamma ray attenuation and radionuclide concentration within an item that can be used as a diagnostic tool. Item preparation is limited to avoiding large quantities of heavily attenuating materials (such as lead shielding) in order to allow sufficient transmission through the container and the matrix.1.1 This test method describes the nondestructive assay (NDA) of gamma ray emitting radionuclides inside containers using tomographic gamma scanning (TGS). High resolution gamma ray spectroscopy is used to detect and quantify the radionuclides of interest. The attenuation of an external gamma ray transmission source is used to correct the measurement of the emission gamma rays from radionuclides to arrive at a quantitative determination of the radionuclides present in the item. 1.2 The TGS technique covered by the test method may be used to assay scrap or waste material in cans or drums in the 1 to 500 litre volume range. Other items may be assayed as well. 1.3 The test method will cover two implementations of the TGS procedure: (1) Isotope Specific Calibration that uses standards of known radionuclide masses (or activities) to determine system response in a mass (or activity) versus corrected count rate calibration, that applies to only those specific radionuclides for which it is calibrated, and (2) Response Curve Calibration that uses gamma ray standards to determine system response as a function of gamma ray energy and thereby establishes calibration for all gamma emitting radionuclides of interest. 1.4 This test method will also include a technique to extend th......

Standard Test Method for Nondestructive Assay of Radioactive Material by Tomographic Gamma Scanning

ASTM C996-10 235U低于5％的浓缩六氟化铀标准规格

1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the production of sinterable UO2 powder for fuel fabrication. In addition to this specification, the parties concerned may agree to other appropriate conditions. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents or requirements for health and safety or for shipping. Observance of this specification does not relieve the user of the obligation to conform to all applicable international, federal, state, and local regulations for processing, shipping, or in any other way using UF6 (see, for example, TID-7016, DP-532, and DOE O474.1). 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Standard Specification for Uranium Hexafluoride Enriched to Less Than 5 % ²³⁵U