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

ASTM C1346-08(2014) P-10管中UF 6溶解的标准操作规程

4.1 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 C787 and C996. 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 C761. 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

ASTM E526-08 用钛的放射性测定快中子反应率的标准试验方法

Refer to Guide E 844 for the selection, irradiation, and quality control of neutron dosimeters. Refer to Test Method E 261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors. Titanium has good physical strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1675°C, and can be obtained with satisfactory purity. 46Sc has a half-life of 83.79 days. The 46Sc decay emits a 0.8893 MeV gamma 99.984 % of the time and a second gamma with an energy of 1.1205 MeV 99.987 % of the time. The isotopic content of natural titanium recommended for 46Ti is 8.25 %. The radioactive products of the neutron reactions 47Ti(n,p)47Sc (τ1/2 = 3.3492 d) and 48Ti(n,p)48Sc (τ1/2 = 43.67 h), might interfere with the analysis of 46Sc. Contaminant activities (for example, 65Zn and 182Ta) might interfere with the analysis of 46Sc. See Sections 7.1.2 and 7.1.3 for more details on the 182Ta and 65Zn interference. 46Ti and 46Sc have cross sections for thermal neutrons of 0.59 and 8 barns, respectively ; therefore, when an irradiation exceeds a thermal-neutron fluence greater than about 2 × 1021 cm–2, provisions should be made to either use a thermal-neutron shield to prevent burn-up of 46Sc or measure the thermal-neutron fluence rate and calculate the burn-up. Fig. 1 shows a plot of cross section versus neutron energy for the fast-neutron reactions of titanium which produce 46Sc [that is, NatTi(n,X)46Sc]. Included in the plot is the 46Ti(n,p) reaction and the 47Ti(n,np) contribution to the 46Sc production, normalized (at 14.7 MeV) per 46Ti atom. This figure is for illustrative purposes only to indicate the range of response of the 46Ti(n,p) reaction. Refer to Guide E 1018 for descriptions of recommended tabulated dosimetry cross sections.1.1 This test method covers procedures for measuring reaction rates by the activation reactions 46Ti(n,p) 46Sc + 47Ti(n, np)46Sc. Note 18212;Since the cross section for the (n,np) reaction is relatively small for energies less than 12 MeV and is not easily distinguished from that of the (n,p) reaction, this test method will refer to the (n,p) reaction only. 1.2 The reaction is useful for measuring neutrons with energies above approximately 4.4 MeV and for irradiation times up to about 250 days (for longer irradiations, see Practice E 261). 1.3 With suitable techniques, fission-neutron fluence rates above 109 cm–2·s–1 can be determined. However, in the presence of a high thermal-neutron fluence rate, 46Sc depletion should be investigated. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261. 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......

Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Titanium

ASTM C1168-08 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 TitleSection Dissolution of Plutonium Metal with Hydrochloric Acid9.1 Dissolution of Plutonium Metal with Sulfuric Acid9.2 Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxide by the Sealed-Reflux Technique9.3 Dissolution of Plutonium Oxide and Uranium-Plutonium Mixed Oxides by Sodium Bisulfate Fusion9.4 Dissolution of Uranium-Plutonium Mixed Oxides and Low-Fired Plutonium Oxide in Beakers9.5 1.3 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 C1347-08(2014)e1 分析用铀材料制备和溶解的标准实施规程

4.1 The materials covered that must meet ASTM specifications are uranium metal and uranium oxide. 4.2 Uranium materials are used as nuclear reactor fuel. For this use, these materials must meet certain criteria for uranium content, uranium-235 enrichment, and impurity content, as described in Specifications C753 and C776. The material is assayed for uranium to determine whether the content is as specified. 4.3 Uranium alloys, refractory uranium materials, and uranium containing scrap and ash are unique uranium materials for which the user must determine the applicability of this practice. In general, these unique uranium materials are dissolved with various acid mixtures or by fusion with various fluxes. 1.1 This practice covers dissolution treatments for uranium materials that are applicable to the test methods used for characterizing these materials for uranium elemental, isotopic, and impurities determinations. Dissolution treatments for the major uranium materials assayed for uranium or analyzed for other components are listed. 1.2 The treatments, in order of presentation, are as follows: Procedure Title Section Dissolution of Uranium Metal and Oxide with Nitric Acid 8.1 Dissolution of Uranium Oxides with Nitric Acid and Residue 8199; Treatment 8.2 Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid 8199;with Residue Treatment 8.3 Dissolution of Uranium Scrap and Ash by Leaching with Nitric 8199;Acid and Treatment of Residue by Carbonate Fusion 8.4 Dissolution of Refractory Uranium-Containing Material by 8199;Carbonate Fusion 8.5 Dissolution of Uranium—Aluminum Alloys Uranium Scrap and Ash, and Refractory Uranium-Containing Materials by Microwave Treatment 8.6 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 Practice for Preparation and Dissolution of Uranium Materials for Analysis

ASTM C751-07 核纯级碳化硼丸的标准规范

1.1 This specification applies to boron carbide pellets for use as a control material in nuclear reactors.1.2 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 Nuclear-Grade Boron Carbide Pellets

ASTM C1205-07(2012) 用α光谱测定法对土壤中镅-241进行放射化学测定的标准试验方法

This test method provides the speed and high decontamination factors attainable with liquid-liquid extraction of the actinides and eliminates filtration techniques that are more time consuming. This test method provides a precise determination of americium in concentrations normally found in environmental samples.1.1 This method covers the determination of americium–241 in soil by means of chemical separations and alpha spectrometry. It is designed to analyze up to ten grams of soil or other sample matrices that contain up to 30 mg of combined rare earths. This method allows the determination of americium–241 concentrations from ambient levels to applicable standards. The values stated in SI units are to be regarded as standard. 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. For specific precaution statements, see Section 10.

Standard Test Method for The Radiochemical Determination of Americium-241 in Soil by Alpha Spectrometry

ASTM C1454-07 支持金属铀废核燃料自燃性分析的自燃性/易燃性测试用标准指南

1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (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 and metallic uranium-based 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 10CFR60/63). 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 predominantly (approximately 80 % by weight of uranium) consists of metallic uranium.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 enhances the chemical reactivity of the SNF by further rupturing the cladding and creating uranium hydride particulates and/or inclusions in the uranium metal matrix. 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 Zircaloy spent fuel cladding has occasionally demonstrated pyrophoric behavior. This behavior often occurred on cladding pieces or particulate residues left after the chemical dissolution of metallic uranium or uranium dioxide during fuel reprocessing of commercial spent fuel and/or extraction of plutonium from defense reactor spent fuel. Although it is generally believed that zirconium is not as intrinsically prone to pyrophoric behavior as uranium or plutonium, it has in the past ignited after being sensitized during the chemical extraction process. Although this guide primarily addresses the pyrophoricity of the metallic uranium component of the spent fuel, some of the general principles involved could also apply to zirconium alloy spent fuel cladding.1.7 The interpretation of the test data de......

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

ASTM C1430-07 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

Uranium dioxide is used as a nuclear-reactor fuel. This test method is designed to determine whether the percent uranium and O/U or O/M content meet Specifications C 776 and C 922.1.1 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 8.

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 C1672-07 用热电离质谱分光计通过总数蒸发法测定铀或钚同位素组成或浓度的标准试验方法

1.1 This method describes the determination of the isotopic composition and/or the concentration of uranium and plutonium as nitrate solutions by the thermal ionization mass spectrometric (TIMS) total evaporation method. Purified uranium or plutonium nitrate solutions are loaded onto a degassed metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion beams are continually measured until the sample is exhausted. The measured ion currents are integrated over the course of the run, and normalized to a reference isotope ion current to yield isotopic ratios.1.2 In principle, the total evaporation method should yield isotopic ratios that do not require mass bias correction. In practice, some samples may require this bias correction. When compared to the conventional TIMS method, the total evaporation method is approximately two times faster, improves precision from two to four fold, and utilizes smaller sample sizes.1.3 The total evaporation method may lead to biases in minor isotope ratios due to peak tailing from adjacent major isotopes, depending on sample characteristics. The use of an electron multiplier equipped with an energy filter may eliminate or diminish peak tailing effects. Measurement of instrument abundance sensitivity may be used to ensure that such biases are negligible, or may be used to bias correct minor isotope ratios.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 Determination of Uranium or Plutonium Isotopic Composition or Concentration by the Total Evaporation Method Using a Thermal Ionization Mass Spectrometer

ASTM E265-07e1 用硫-32的放射性测量快速中子流量密度和反应速率的测试方法

Refer to Guides E 720 and E 844 for the selection, irradiation, and quality control of neutron dosimeters. Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence and fluence rate with threshold detectors. The activation reaction produces 32P, which decays by the emission of a single beta particle in 100 % of the decays, and which emits no gamma rays. The half life of 32P is 14.262 (14) days (1) and the maximum beta energy is 1710 keV(2). Elemental sulfur is readily available in pure form and any trace contaminants present do not produce significant amounts of radioactivity. Natural sulfur, however, is composed of 32S (95.02 % (9)), 34S (4.21 % (8)) (1), and trace amounts of other sulfur isotopes. The presence of these other isotopes leads to several competing reactions that can interfere with the counting of the 1710-keV beta particle. This interference can usually be eliminated by the use of appropriate techniques, as discussed in Section 8.1.1 This test method describes procedures for measuring reaction rates and fast-neutron fluences by the activation reaction 32S(n,p)32P. 1.2 This activation reaction is useful for measuring neutrons with energies above approximately 3 MeV. 1.3 With suitable techniques, fission-neutron fluences from about 5 × 108 to 1016 n/cm2 can be measured. 1.4 Detailed procedures for other fast-neutron detectors are described in Practice E 261. 1.5 This standard does not purport to address all of the safety problems, 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 Measuring Reaction Rates and Fast-Neutron Fluences by Radioactivation of Sulfur-32

ASTM E385-07 使用14兆电子伏特的中子活化和直接计数技术测定氧含量的标准试验方法

The conventional determination of oxygen content in liquid or solid samples is a relatively difficult chemical procedure. It is slow and usually of limited sensitivity. The 14-MeV neutron activation and direct counting technique provides a rapid, highly sensitive, nondestructive procedure for oxygen determination in a wide range of matrices. This test method is independent of the chemical form of the oxygen. This test method can be used for quality and process control in the metals, coal, and petroleum industries, and for research purposes in a broad spectrum of applications.1.1 This test method covers the measurement of oxygen concentration in almost any matrix by using a 14-MeV neutron activation and direct-counting technique. Essentially, the same system may be used to determine oxygen concentrations ranging from over 50 % to about 10 g/g, or less, depending on the sample size and available 14-MeV neutron fluence rates. Note 1 - The range of analysis may be extended by using higher neutron fluence rates, larger samples, and higher counting efficiency detectors.1.2 This test method may be used on either solid or liquid samples, provided that they can be made to conform in size, shape, and macroscopic density during irradiation and counting to a standard sample of known oxygen content. Several variants of this method have been described in the technical literature. A monograph is available which provides a comprehensive description of the principles of activation analysis using a neutron generator (1).1.3 The values stated in either SI or inch-pound units are to be regarded separately as the standard. The values given in parentheses are for information only.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 precautions are given in Section 8.

Standard Test Method for Oxygen Content Using a 14-MeV Neutron Activation and Direct-Counting Technique

ASTM C1455-07 用Υ射线光谱法对特殊核物质含量无损分析的标准试验方法

1.1 This test method describes gamma-ray methods used to nondestructively measure the quantity of 235U, or 239Pu remaining as holdup in nuclear facilities. Holdup occurs in all facilities where nuclear material is processed, in process equipment, in exhaust ventilation systems and in building walls and floors.1.2 This test method includes information useful for management, planning, selection of equipment, consideration of interferences, measurement program definition, and the utilization of resources (1, 2, 3, 4).1.3 The measurement of nuclear material hold up in process equipment requires a scientific knowledge of radiation sources and detectors, transmission of radiation, calibration, facility operations and error analysis. It is subject to the constraints of the facility, management, budget, and schedule; plus health and safety requirements; as well as the laws of physics. The measurement process includes defining measurement uncertainties and is sensitive to the form and distribution of the material, various backgrounds, and interferences. The work includes investigation of material distributions within a facility, which could include potentially large holdup surface areas. Nuclear material held up in pipes, ductwork, gloveboxes, and heavy equipment, is usually distributed in a diffuse and irregular manner. It is difficult to define the measurement geometry, to identify the form of the material, and to measure it without interference from adjacent sources of radiation.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 Special Nuclear Material Holdup Using Gamma-Ray Spectroscopic Methods

ASTM C1430-07(2011)e1 用大气平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

Uranium dioxide is used as a nuclear-reactor fuel. This test method is designed to determine whether the percent uranium and O/U or O/M content meet Specifications C776 and C922.1.1 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. 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 hazards statements, see Section 8.

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 C1672-07(2014) 用热电离质谱分光计通过总数蒸发法测定铀或钚同位素组成或浓度的标准试验方法

5.1 The total evaporation method is used to measure the isotopic composition of uranium and plutonium materials, and may be used to measure the elemental concentrations of the two elements when employing the IDMS technique. 5.2 Uranium and plutonium compounds are used as nuclear reactor fuels. In order to be suitable for use as a nuclear fuel the starting material must meet certain specifications, such as found in Specifications C757, C833, C753, C776, C787, C967, C996, C1008, or as specified by the purchaser. The uranium and/or plutonium concentration and isotopic abundances are measured by mass spectrometry following this method. 5.3 The total evaporation method allows for a wide range of sample loading with no loss in precision or accuracy, and is also suitable for trace-level loadings with consequent loss of precision. Typical uranium analyses are conducted using sample loadings between 10 nanograms and several micrograms. Plutonium analyses are generally conducted using between five and 200 nanograms of plutonium per filament. The total evaporation method and modern instrumentation allow for the measurement of minor isotopes using ion counting detectors, while the major isotopes are simultaneously measured using Faraday cup detectors. 5.4 New generations of miniaturized ion counters now allow extremely small samples, in the picogram to femtogram range, to be measured via total evaporation methods. The method may be employed for measuring environmental or safeguards inspection samples containing very small quantities of uranium or plutonium. Very small loadings require special sample handling and analysis techniques, and careful evaluation of measurement uncertainty contributors. 1.1 This method describes the determination of the isotopic composition and/or the concentration of uranium and plutonium as nitrate solutions by the thermal ionization mass spectrometric (TIMS) total evaporation method. Purified uranium or plutonium nitrate solutions are loaded onto a degassed metal filament and placed in the mass spectrometer. Under computer control, ion currents are generated by heating of the filament(s). The ion beams are continually measured until the sample is exhausted. The measured ion currents are integrated over the course of the run, and normalized to a reference isotope ion current to yield isotopic ratios. 1.2 In principle, the total evaporation method should yield isotopic ratios that do not require mass bias correction. In practice, some samples may require this bias correction. When compared to the conventional TIMS method, the total evaporation method is approximately two times faster, improves precision from two to four fold, and utilizes smaller sample sizes. 1.3 The total evaporation method may lead to biases in minor isotope ratios due to peak tailing from adjacent major isotopes, depending on sample characteristics. The use of an electron multiplier equipped with an energy filter may eliminate or diminish peak tailing effects. Measurement of instrument abundance sensitivity may be used to ensure that such biases are negligible, or may be used to bias correct minor isotope ratios. 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 Test Method for Determination of Uranium or Plutonium Isotopic Composition or Concentration by the Total Evaporation Method Using a Thermal Ionization Mass Spectrometer

ASTM C1205-07 用α光谱测定法对土壤中镅-241进行放射化学测定的标准试验方法

This test method provides the speed and high decontamination factors attainable with liquid-liquid extraction of the actinides and eliminates filtration techniques that are more time consuming. This test method provides a precise determination of americium in concentrations normally found in environmental samples.1.1 This method covers the determination of americium-241 in soil by means of chemical separations and alpha spectrometry. It is designed to analyze up to ten grams of soil or other sample matrices that contain up to 30 mg of combined rare earths. This method allows the determination of americium-241 concentrations from ambient levels to applicable standards. The values stated in SI units are to be regarded as standard.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 precaution statements, see Section 10.

Standard Test Method for The Radiochemical Determination of Americium-241 in Soil by Alpha Spectrometry

ASTM C1477-06 多收集器感应耦合等离子质谱法分析六氟化铀和硝酸铀酰溶液同位素丰度的标准试验方法

The test method is capable of measuring uranium isotopic abundances of 234U, 235U, 236U and 238U as required by Specifications C 787 and C 996.1.1 This test method covers the isotopic abundance analysis of 234U, 235U, 236U and 238U in samples of hydrolysed uranium hexafluoride (UF6) by inductively coupled plasma source, multi-collector, mass spectrometry (ICP-MC-MS). The method applies to material with 235U abundance in the range of 0.2 to 6 % mass. This test method is also described in ASTM STP 1344.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 Hexafluoride and Uranyl Nitrate Solutions by Multi-Collector, Inductively Coupled Plasma-Mass Spectrometry

ASTM C757-06 核纯级可烧结的二氧化钚粉末的规格

1.1 This specification covers nuclear grade plutonium dioxide, sinterable powder obtained by the oxalate precipitation route, calcined above 500C, or any other equivalent process acceptable to the buyer. Included is plutonium dioxide of various isotopic compositions as normally prepared by in-reactor neutron irradiation of natural or slightly enriched uranium or by in-reactor neutron irradiation of recycled plutonium mixed with uranium.1.2 There is no discussion of or provision for preventing criticality incidents, nor are health and safety requirements, the avoidance of hazards, or shipping precautions and controls discussed. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all national and local regulations on processing, shipping, or using source or special nuclear materials. For examples in the U.S. Government, relevant documents are Code of Federal Regulations, Title 10 Nuclear Safety Guide, U.S. Atomic Energy Commission Report TID-7016, and "Handbook of Nuclear Safety", H. K. Clark, U.S. Atomic Energy Commission Report, DP-532.1.3 The PuO2 shall be produced by a qualified process and in accordance with a quality assurance program approved by the user.1.4 The values stated in SI units are to be regarded as the standard.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 Nuclear-Grade Plutonium Dioxide Powder, Sinterable

ASTM C968-06 烧结氧化钆及二氧化铀丸粒的标准试验方法

The test methods in this method are designed to show whether a given material is in accordance with Specification C 922.1.1 These test methods cover procedures for the analysis of sintered gadolinium oxide-uranium dioxide pellets to determine compliance with specifications.1.2 The analytical procedures appear in the following order:1.3 The values stated in SI units are to be regarded as the standard.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 This test method describes the determination of nitrogen in gadolinium oxide-uranium dioxide pellets (Gd2O 3/UO2). With a 2 to 5-g sample, concentrations from 5 to 100 g of nitrogen are determined without interference.1.5 The homogeneity of Gd2O3 in UO2 has been cited in Specification C 922 as an important requirement for this fuel form. The uniform distribution of gadolinia in urania will result in up to three components in the pellet: free Gd 2O3, free UO2, and a Gd2O3-UO 2 solid solution. There are a number of ways for assessing uniformity of which the ceramographic method described here may not be the most definitive. This technique has been used over the gadolinia concentration range from 1 to 10 weight %.

Standard Test Methods for Analysis of Sintered Gadolinium Oxide-Uranium Dioxide Pellets

ASTM C787-06 浓缩用六氟化铀标准规范

1.1 This specification covers uranium hexafluoride (UF6) intended for feeding to an enrichment plant. Included are specifications for UF6 derived from unirradiated natural uranium and UF6 derived from irradiated uranium that has been reprocessed and converted to UF6 for enrichment and subsequent reuse. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Commercial Natural UF6 feedstock so that the corresponding enriched uranium is essentially equivalent to enriched uranium made entirely from virgin natural UF6; and ( 2) To define additional limits for Reprocessed UF6 (or any mixture of Reprocessed UF6 and Commercial Natural UF6). For such UF6, special provisions may be needed to ensure that no extra hazard arises to the work force, process equipment, or the environment.1.2 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 international, federal, state, and local regulations for processing, shipping, or in any other way using UF6 (see, for example, TID-7016, DP-532, ORNL-NUREG-CSD-6, and DOE O 474.1).1.3 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 for Enrichment

ASTM C1636-06a 六氟化铀中铀232测定的标准指南

The method is applicable to the analysis of materials to demonstrate compliance with the specifications set forth in Specifications C 787 and C 996.1.1 This method covers the determination of 232U in uranium hexafluoride by alpha spectrometry.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 to determine the applicability of regulatory limitations prior to use.

Standard Guide for the Determination of Uranium-232 in Uranium Hexafluoride