27.120.30 (Fissile materials and nuclear fuel tech 标准查询与下载

ASTM C1168-01 分析用钚材料的制备和溶解的标准实施规程

The materials covered are plutonium metal, plutonium oxide, and uranium-plutonium mixed oxide, including those that must meet ASTM product specifications. Plutonium and uranium mixtures are used as nuclear reactor fuels. For use as a nuclear reactor fuel, the material must meet certain criteria for combined uranium and pluto- nium content, effective fissile content, and impurity content as described in Specifications C 757, C 833, and C 1008. The material is assayed for plutonium and uranium to determine if the content is correct as specified by the purchaser. The materials not covered are unique plutonium materials, including alloys, compounds, and scrap materials. The user must determine the applicability of this practice to these other materials. In general, these unique plutonium materials are dissolved with various acid mixtures or by fusion with various fluxes. Many plutonium salts are soluble in hydrochloric acid.1.1 This practice is a compilation of dissolution techniques for plutonium materials that are applicable to the test methods used for characterizing these materials. Dissolution treatments for the major plutonium materials assayed for plutonium or analyzed for other components are listed. Aliquants of the dissolved samples are dispensed on a weight basis when one of the analyses must be highly reliable, such as plutonium assay; otherwise they are dispensed on a volume basis.1.2 The treatments, in order of presentation, are as follows:Procedure TitleSectionDissolution of Plutonium Metal with Hydrochloric Acid7.1Dissolution of Plutonium Metal with Sulfuric Acid7.2Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxide by the Sealed-Reflux Technique7.3Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion7.4Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers7.51.3 The values stated in SI units are to be regarded as 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.

Standard Practice for Preparation and Dissolution of Plutonium Materials for Analysis

ASTM C1052-01 液体六氟化铀散装取样标准实施规程

1.1 This practice covers methods for withdrawing representative samples of liquid uranium hexafluoride (UF6) from bulk quantities of the material. Such samples are used for determining compliance with the applicable commercial specification, for example Specification C 787 and Specification C 996.1.2 It is assumed that the bulk liquid UF6 being sampled comprises a single quality and quantity of material. This practice does not address any special additional arrangements that might be required for taking proportional or composite samples, or when the sampled bulk material is being added to UF6 residues already in a container ("heels recycle").1.3 The number of samples to be taken, their nominal sample weight, and their disposition shall be agreed upon between the parties.1.4 The scope of this practice does not include provisions for preventing criticality incidents.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 Practice for Bulk Sampling of Liquid Uranium Hexafluoride

ASTM C833-01 烧结的二氧化铀、钚芯块标准规范

1.1 This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. This specification may not completely cover the requirements for pellets fabricated from weapons-derived plutonium.1.2 This specification does not include ( 1) provisions for preventing criticality accidents or ( 2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50-Domestic Licensing of Production and Utilization Facilities; Title 10, Part 71-Packaging and Transportation of Radioactive Material; and Title 49, Part 173-General Requirements for Shipments and Packaging.1.3 The following safety hazards caveat pertains only to the technical requirements portion, Section 4, of this specification: 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.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 Sintered (Uranium-Plutonium) Dioxide Pellets

ASTM C1462-00(2005) 大于15%小于20%的U-235的浓缩铀金属的标准规范

1.1 This specification covers nuclear grade uranium metal that has either 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 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section . The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal.1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. 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 standard does not relieve the user of the obligation to conform to all applicable international, federal, state, and local regulations for processing, shipping, or any other way of using uranium metal (see, for example, C 996 regarding references).

Standard Specification for Uranium Metal Enriched to More than 15% and Less Than 20%²³⁵ U

ASTM C1453-00 用点火和核级二氧化铀粉末和颗粒的氧铀原子比测定铀的标准试验方法

1.1 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear grade uranium dioxide powder and pellets.1.2 This test method does not include provisions for preventing criticality accidents or requirements for health and safety. Observance of this test method does not relieve the user of the obligation to be aware of and conform to all international, national, or federal, state and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material.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.1.3 This test method also is applicable to UO 3 and U3O8 powder.

Standard Test Method for the Determination of Uranium by Ignition and the Oxygen to Uranium (O/U) Atomic Ratio of Nuclear Grade Uranium Dioxide Powders and Pellets

ASTM C1476-00 感应耦合等离子质谱法分析锝99用尿的标准试验方法

1.1 This test method covers the determination of the concentration of technetium-99 in urine using inductively coupled plasma-mass spectrometry (ICP-MS). This test method can be used to support uranium enrichment and reclamation facility bioassay programs.1.2 The minimum detectable concentration for this test method, using a quadrupole ICP-MS, is approximately 1.0 ng/L (0.62 Bq/L).

Standard Test Method for Analysis of Urin for Technetium-99 by Inductively Coupled Plasma-Mass Spectrometry

ASTM C922-00 烧结氧化钆及二氧化铀丸标准规范

1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unreprocessed uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all international, federal, state, and local regulations on possessing, shipping, processing, or using source or special nuclear material. Examples of U.S. Governmental documents are Code of Federal Regulations (Latest Edition), Title 10, Part 50, Title 10, Part 71, and Title 49, Part 173. 1.4 The following precautionary caveat pertains only to the technical requirements portion, Section 4, of this specification: 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 Specification for Sintered Gadolinium Oxide-Uranium Dioxide Pellets

ASTM E692-00 用高分辨率γ射线光谱分析法对照射过的核燃料中的铯137的标准试验方法

This test method uses a high-resolution gamma-ray spectrometer as a basis for measuring the gamma-ray emission rate of 137Cs-137mBa in a dilute nitric acid solution containing 10 mg/L of cesium carrier. No chemical separation of the cesium from the dissolved-fuel solution is required. The principal steps consist of diluting a weighed aliquot of the dissolved-fuel solution with a known mass of 1 M nitric acid (HNO3) and measuring the 662 keV gamma-ray count rate from the sample, then measuring the 662 keV gamma-ray count rate from a standard source that has the same physical form and counting geometry as the sample. The amount of fuel sample required for the analysis is small. For a sample containing 1 mg of fuel irradiated to one atom percent fission, a net count rate of approximately 103 counts per second will be observed for a counting geometry that yields a full-energy peak efficiency fraction of 1 × 10-3. The advantage of this small amount of sample is that the concentration of fuel material can be kept at levels well below 1 g/L, which results in negligible self-absorption in the sample aliquot and a small radiation hazard to the analyst.1.1 This test method covers the determination of the number of atoms of Cs in aqueous solutions of irradiated uranium and plutonium nuclear fuel. To minimize interference from short-lived fission products, the fuel should decay 4 months or more prior to the gamma-ray emission-rate measurement. 1.2 When combined with a method for determining the initial number of fissile atoms in the fuel, the results of this analysis allows atom percent fission (burnup) to be calculated. The determination of atom percent fission, uranium and plutonium concentrations, and isotopic abundances are covered in Test Methods E267 and E321. 1.3 Cs is not suitable as a fission monitor for samples that may have lost cesium during reactor operation. For example, a large temperature gradient enhances Cs migration from the fuel region to cooler regions such as the radial fuel-clad gap, or to a lesser extent, towards the axial fuel end. 1.4 The Cs distribution may be ascertained by an axial gamma-ray scan of the fuel element to be assayed. In a mixed-oxide fuel, comparison of the Cs distribution with the distribution of nonmigrating fission-product nuclides such as Zr or Ce would indicate the relative degree of Cs migration. A nonuniform Cs distribution should alert the analyst to the potential loss of the fission nuclide. 1.5 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety problems associated with its use. It is the responsibility of whoever uses this standard to consult and establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Determining the Content of Cesium-137 in Irradiated Nuclear Fuels by High-Resolution Gamma-Ray Spectral Analysis

ASTM C1477-00 多收集器感应耦合等离子质谱法分析六氟化铀同位素丰度的标准试验方法

1.1 This test method covers the isotopic abundance analysis or 234U, 235U and 238U in samples of hydrolysed uranium hexafluoride (UF6) by inductively coupled plasma source, multi-collector, mass spectrometry (ICP-MC-MS). This test method is also describe in ASTM STP 1344. 1.2 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 Abundance Analysis of Uranium Hexafloride by Multi-Collector, Inductively Coupled Plasma-Mass Spectrometry

ASTM C1462-00(2008) 大于15%小于20%的U-235的浓缩铀金属的标准规范

1.1 This specification covers nuclear grade uranium metal that has either 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 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. 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 standard does not relieve the user of the obligation to conform to all applicable international, federal, state, and local regulations for processing, shipping, or any other way of using uranium metal (see, for example, C 996 regarding references).

Standard Specification for Uranium Metal Enriched to More than 15 % and Less Than 20 % ²³⁵U

ASTM C1454-00 通过对燃烧核燃料的金属铀的自燃分析进行自燃/易燃性试验的标准指南

1.1 This guide describes testing protocols for pyrophoricity, or combustibility characteristics, or both, of metallic uranium-based SNF. The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI).1.2 This guide describes testing of metallic uranium spent nuclear fuel (SNF) in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60). The testing described herein is designed to provide basic data related to the evaluation of the pyrophoricity/combustibility characteristics of containers or waste packages containing metallic uranium SNF in support of safety analyses (SAR), or performance assessments (PA) of transport, storage, or disposal systems, or a combination thereof.1.3 Spent nuclear fuel that is not reprocessed must be emplaced in secure temporary interim storage as a step towards its final disposal in a geologic repository. In the United States, SNF, from both civilian commercial power reactors and defense nuclear materials production reactors, will be sent to interim storage, and subsequently, to deep geologic disposal. U.S. commercial SNF comes predominantly from light water reactors (LWRs) and is uranium dioxide-based, whereas U.S. Department of Energy (DOE) owned defense reactor SNF is in several different chemical forms, but is predominantly (80 % by weight of uranium) metallic uranium-based.1.4 Knowledge of the pyrophoricity/combustibility characteristics of the SNF is required to support licensing activities for extended interim storage and ultimate disposition in a geologic repository. These activities could include interim storage configuration safety analyses, conditioning treatment development, preclosure design basis event (DBE) analyses of the repository controlled area, and postclosure performance assessment of the EBS.1.5 Metallic uranium fuels are clad, generally with zirconium, aluminum, stainless steel, or magnesium alloy, to prevent corrosion of the fuel and to contain fission products. If the cladding is damaged and the metallic SNF is stored in water the consequent corrosion and swelling of the exposed uranium may enhance the chemical reactivity of the SNF by further rupturing the cladding and creating uranium hydride particulates and/or inclusions. The condition of the metallic SNF will affect its behavior in transport, interim storage or repository emplacement, or both, and therefore, influence the engineering decisions in designing the pathway to disposal.1.6 The interpretation of the test data depends on the characteristics of the sample tested. The type and the size of the SNF sample must be chosen carefully and accounted for in the usage of the data. The use of the data obtained by the testing described herein may require that samples be used which mimic the condition of the SNF at times far into the future, for example, the repository postcontainment period. This guide does not specifically address methods for `aging'' samples for this purpose. The section in Practice C 1174 concerning the accelerated testing of waste package materials is recommended for guidance on this subject.

Standard Guide for Pyrophoricity/Combustibility Testing in Support of Pyrophoricity Analyses of Metallic Uranium Spent Nuclear Fuel

ASTM C1062-00(2008) 核燃料溶解装置的设计、制造和安装的标准指南

The purpose of this guide is to provide information that will help to ensure that nuclear fuel dissolution facilities are conceived, designed, fabricated, constructed, and installed in an economic and efficient manner. This guide will help facilities meet the intended performance functions, eliminate or minimize the possibility of nuclear criticality and provide for the protection of both the operator personnel and the public at large under normal and abnormal (emergency) operating conditions as well as under credible failure or accident conditions.1.1 It is the intent of this guide to set forth criteria and procedures for the design, fabrication and installation of nuclear fuel dissolution facilities. This guide applies to and encompasses all processing steps or operations beyond the fuel shearing operation (not covered), up to and including the dissolving accountability vessel. 1.2 Applicability and Exclusions: 1.2.1 Operations8212;This guide does not cover the operation of nuclear fuel dissolution facilities. Some operating considerations are noted to the extent that these impact upon or influence design. 1.2.1.1 Dissolution Procedures8212;Fuel compositions, fuel element geometry, and fuel manufacturing methods are subject to continuous change in response to the demands of new reactor designs and requirements. These changes preclude the inclusion of design considerations for dissolvers suitable for the processing of all possible fuel types. This guide will only address equipment associated with dissolution cycles for those fuels that have been used most extensively in reactors as of the time of issue (or revision) of this guide. (See Appendix X1.) 1.2.2 Processes8212;This guide covers the design, fabrication and installation of nuclear fuel dissolution facilities for fuels of the type currently used in Pressurized Water Reactors (PWR). Boiling Water Reactors (BWR), Pressurized Heavy Water Reactors (PHWR) and Heavy Water Reactors (HWR) and the fuel dissolution processing technologies discussed herein. However, much of the information and criteria presented may be applicable to the equipment for other dissolution processes such as for enriched uranium-aluminum fuels from typical research reactors, as well as for dissolution processes for some thorium and plutonium-containing fuels and others. The guide does not address equipment design for the dissolution of high burn-up or mixed oxide fuels. 1.2.2.1 This guide does not address special dissolution processes that may require substantially different equipment or pose different hazards than those associated with the fuel types noted above. Examples of precluded cases are electrolytic dissolution and sodium-bonded fuels processing. The guide does not address the design and fabrication of continuous dissolvers. 1.2.3 Ancillary or auxiliary facilities (for example, steam, cooling water, electrical services) are not covered. Cold chemical feed considerations are addressed briefly. 1.2.4 Dissolution Pretreatment8212;Fuel pretreatment steps incidental to the preparation of spent fuel assemblies for dissolution reprocessing are not covered by this guide. This exclusion applies to thermal treatment steps such as “Voloxidation” to drive off gases prior to dissolution, to mechanical decladding operations or process steps associated with fuel elements disassembly and removal of end fittings, to chopping and shearing operations, and to any other pretreatment operations......

Standard Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities

ASTM C1458-00 用量热分析进行钚、氚和241Am无损试验的标准试验方法

1.1 This test method describes the nondestructive assay (NDA) of plutonium, tritium, and 241Am using heat flow calorimetry. For plutonium the range of applicability corresponds to 1 g to > 2000 g quantities while for tritium the range extends from 0.001 g to > 10 g. This test method can be applied to materials in a wide range of container sizes up to 50 L. It has been used routinely to assay items whose thermal power ranges from 0.001 W to 135 W.1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes and the 241Am/Pu mass ratio to determine the total plutonium mass.1.3 This test method provides a direct measure of tritium content.1.4 This test method provides a measure of 241Am either as a single isotope or mixed with plutonium.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 Plutonium, Tritium and ²⁴¹Am by Calorimetric Assay

ASTM C1453-00(2006) 用点火和核级二氧化铀粉末和颗粒的氧铀原子比测定铀的标准试验方法

The test method is designed to show whether or not a material meets the specifications as given in Specifications C 753 or C 776. The powderrsquo;stoichiometry is useful for predicting the oxidersquo;sintering behavior in the pellet production process.1.1 This test method covers the determination of uranium and the oxygen to uranium atomic ratio in nuclear grade uranium dioxide powder and pellets.1.2 This test method does not include provisions for preventing criticality accidents or requirements for health and safety. Observance of this test method does not relieve the user of the obligation to be aware of and conform to all international, national, or federal, state and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material.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.1.3 This test method also is applicable to UO 3 and U3O8 powder.

Standard Test Method for the Determination of Uranium by Ignition and the Oxygen to Uranium (O/U) Atomic Ratio of Nuclear Grade Uranium Dioxide Powders and Pellets

ASTM C1474-00 四极感应耦合等离子质谱法分析核纯级燃料中铀的同位素组分的标准试验方法

1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described in ASTM STP 1344.1.2 The 233U isotope is primarily measured as a qualitative measure of its presence by comparing the 233U peak intensity to a background point since it is not normally found present in materials. The example data presented in this test method do not contain any 233U data. A 233U enriched standard is given in Section , and it may be used as a quantitative spike addition to the other standard materials listed.1.3 A single standard calibration technique is used. Optimal accuracy (or a low bias) is achieved through the use of a single standard that is closely matched to the enrichment of the samples. The intensity or concentration is also adjusted to within a certain tolerance range to provide good statistical counting precision for the low-abundance isotopes while maintaining a low bias for the high-abundance isotopes, resulting from high-intensity dead time effects. No blank subtraction or background correction is utilized. Depending upon the standards chosen, enrichments between depleted and 97 % can be quantified. The calibration and measurements are made by measuring the intensity ratios of each low-abundance isotope to the intensity sum of 233U, 234U, 235U, 236U, and 238U. The high-abundance isotope is obtained by difference.1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. The instrument is calibrated and the samples measured in units of isotopic weight percent (Wt %). For example, the 235U enrichment may be stated as Wt % 235U or as g 235U/100 g of U. Statements regarding dilutions, particularly for ug/g concentrations or lower, are given assuming a solution density of 1.0 since the uranium concentration of a solution is not important when making isotopic ratio measurements other than to maintain a reasonably consistent intensity within a tolerance range.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 precautionary statements are given in Section 9.

Standard Test Method for Analysis of Isotopic Composition of Uranium in Nuclear-Grade Fuel Material by Quadrupole Inductively Coupled Plasma-Mass Spectrometry

ASTM C1455-00 用Υ射线光谱法对特殊核物质含量的无损分析的标准指南

1.1 This guide addresses methods used to prepare for and to perform, using gamma-ray measurements, the nondestructive assay (NDA) of radioisotopes, for example, 235U, or 239Pu, remaining as holdup in nuclear facilities. Holdup occurs in facilities where nuclear material is processed. This guide includes the measurement of holdup of Special Nuclear Material (SNM) in places where holdup may occur, such as in process equipment, and in exhaust ventilation systems. This guide includes information useful for management planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources.1.2 The measurement of nuclear material help up in process equipment is both an art and a science. It is subject to the constraints of politics, economics plus health and safety requirements, as well as to the laws of physics. The measurement process often is long and tedious and is performed under difficult circumstances of location and environment. The work combines the features of a detective investigation and a treasure hunt. Nuclear material held up in pipes, ductwork, gloveboxes, heavy equipment, and so forth, usually is distributed in a diffuse and irregular manner. It is difficult to define the measurement geometry, identify the form of the material, and measure it without interference from adjacent sources of radiation. A scientific knowledge of radiation sources and detectors, calibration procedures, geometry and error analysis also is needed ().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 Nondestructive Assay of Special Nuclear Material Holdup Using Gamma-Ray Spectroscopic Methods

ASTM C1430-00 用大气均衡法测定烧结的二氧化铀和氧化钆-二氧化铀颗粒中铀、氧铀比(O/U)和氧金属比(O/M)的标准试验方法

1.1 This test method applies to the determination of uranium, the oxygen to uranium (O/U) ratio in sintered uranium dioxide pellets, and the oxygen to metal (O/M) ratio in sintered gadolinium oxide-uranium dioxide pellets with a Gd2O3 concentration of up to 12 weight %. The O/M calculations assume that the gadolinium and uranium oxides are present in a metal dioxide solid solution.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 statements, see Section .

Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

ASTM C1001-00 用α光谱分析法放化测定土壤中钚的标准试验方法

1.1 This test method covers the determination of plutonium in soils at levels of detection dependent on count time, sample size, detector efficiency, background, and tracer yield. This test method describes one acceptable approach to the determination of plutonium in soil. 1.2 This test method is designed for 10 g of soil, previously collected and treated as described in Practices C998 and C999, but sample sizes up to 50 g may be analyzed by this test method. This test method may not be able to completely dissolve all forms of plutonium in the soil matrix. 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. Specific hazard statements are given in Sections 6 and 9.

Standard Test Method for Radiochemical Determination of Plutonium in Soil by Alpha Spectroscopy

ASTM C1457-00(2005) 用载气萃取法测定二氧化铀粉末或颗粒中氢的总含量的标准试验方法

Uranium dioxide is used as a nuclear-reactor fuel. Gadolinium oxide is used as an additive to uranium dioxide. In order to be suitable for this purpose, these materials must meet certain criteria for impurity content. This test method is designed to determine whether the hydrogen content meets Specifications C 753, C 776, C 888, and C 922.1.1 This test method applies to the determination of hydrogen in nuclear-grade uranium oxide powders and pellets to determine compliance with specifications. Gadolinium oxide (Gd2O3) and gadolinium oxide-uranium oxide powders and pellets may also be analyzed using this test method.1.2 This standard describes a procedure for measuring the total hydrogen content of uranium oxides. The total hydrogen content results from absorbed water, water of crystallization, hydro-carbides and other hydrogenated compounds which may exist as fuel's impurities.1.3 This test method covers the determination of 0.05 to 200 g of residual hydrogen.1.4 This test method describes an electrode furnace carrier gas combustion system equipped with a thermal conductivity detector.1.5 The preferred system of units is micrograms hydrogen per gram of sample (g/g sample) or micrograms hydrogen per gram of uranium (g/g U).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 the Determination of Total Hydrogen Content of Uranium Oxide Powders and Pellets by Carrier Gas Extraction

ASTM C1474-00(2011) 四极感应耦合等离子质谱法分析核纯级燃料中铀的同位素组分的标准验方法

Nuclear-grade reactor fuel material must meet certain criteria, such as those described in Specifications C 753, C 776, C 778, and C 833. Included in these criteria is the uranium isotopic composition. This test method is designed to demonstrate whether or not a given material meets an isotopic requirement and whether the effective fissile content is in compliance with the purchaserrsquo;specifications.1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described in ASTM STP 1344 . 1.2 The 233U isotope is primarily measured as a qualitative measure of its presence by comparing the 233U peak intensity to a background point since it is not normally found present in materials. The example data presented in this test method do not contain any 233U data. A 233U enriched standard is given in Section 8, and it may be used as a quantitative spike addition to the other standard materials listed. 1.3 A single standard calibration technique is used. Optimal accuracy (or a low bias) is achieved through the use of a single standard that is closely matched to the enrichment of the samples. The intensity or concentration is also adjusted to within a certain tolerance range to provide good statistical counting precision for the low-abundance isotopes while maintaining a low bias for the high-abundance isotopes, resulting from high-intensity dead time effects. No blank subtraction or background correction is utilized. Depending upon the standards chosen, enrichments between depleted and 97 % can be quantified. The calibration and measurements are made by measuring the intensity ratios of each low-abundance isotope to the intensity sum of 233U, 234U, 235U, 236U, and 238U. The high-abundance isotope is obtained by difference. 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. The instrument is calibrated and the samples measured in units of isotopic weight percent (Wt %). For example, the 235U enrichment may be stated as Wt % 235U or as g 235U/100 g of U. Statements regarding dilutions, particularly for ug/g concentrations or lower, are given assuming a solution density of 1.0 since the uranium concentration of a solution is not important when making isotopic ratio measurements other than to maintain a reasonably consistent intensity within a tolerance range. 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 appli......

Standard Test Method for Analysis of Isotopic Composition of Uranium in Nuclear-Grade Fuel Material by Quadrupole Inductively Coupled Plasma-Mass Spectrometry