F40 核材料、核燃料综合 标准查询与下载

ANSI N14.1-2001 核原料.六氟化铀.运输包装

Provides criteria for packaging of uranium hexafluoride (UF6) for transport. This standard also defines the requirements for in-service inspections, cleanliness, and maintenance for packagings in service. Also included are cylinder loadings shipping deta

Nuclear Materials – Uranium Hexafluoride – Packaging for Transport

ASTM C1493-01 用差动衰减系统钝化-活化中子计数法对废物中的核物质进行无损分析的标准试验方法

1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential die-away technique (DDT), and passive neutron coincidence counting. Results from the active and passive measurements are combined to determine the total amount of fissile and spontaneously-fissioning material in drums of scrap or waste as large as 208 L. Corrections are made to the measurements for the effects of neutron moderation and absorption, assuming that the effects are averaged over the volume of the drum and that no significant lumps of nuclear material are present. These systems are most widely used to assay low-level and transuranic waste, but may also be used for the measurement of scrap materials. While this test method is specific to the second-generation Los Alamos National Laboratory (LANL) passive-active neutron assay system, the principle applies to other DDT systems.1.1.1 In the active mode, the system measures fissile isotopes such as 235U and 239Pu. The neutrons from a pulsed, 14-MeV neutron generator are thermalized to induce fission in the assay item. Between generator pulses, the system detects prompt-fission neutrons emitted from the fissile material. The number of detected neutrons between pulses is proportional to the mass of fissile material. This method is called the differential die-away technique.1.1.2 In the passive mode, the system detects time-coincident neutrons emitted from spontaneously fissioning isotopes. The primary isotopes measured are 238Pu, 240Pu, and 242Pu; however, the system may be adapted for use on other spontaneously-fissioning isotopes as well. The number of coincident neutrons detected is proportional to the mass of spontaneously-fissioning material.1.2 The active mode is used to assay fissile material in the following ranges.1.2.1 For uranium-bearing items, the DDT can measure the 235U content in the range from 0.02 to over 100 g. Normally, the assay of items bearing only uranium is performed using matrix-specific calibrations to account for the effect of the matrix on the active signal.1.2.2 For plutonium-bearing items, the DDT method measures the 239Pu content in the range between 0.01 and 20 g.1.3 The passive mode is capable of assaying spontaneously-fissioning nuclei, over a nominal range from 0.05 to 15 g of 240Pu, or equivalent. The passive mode can also be used to measure large (for example, kg) quantities of 238U.1.4 This test method requires knowledge of the relative abundances of the plutonium or uranium isotopes to determine the total plutonium or uranium mass.1.5 This test method will give biased results when the waste form does not meet the calibration specifications and the measurement assumptions presented in this test method regarding the requirements for a homogeneous matrix, uniform source distribution, and the absence of nuclear material lumps, to the extent that they effect the measurement.1.6 The complete active and passive assay of a 208 L drum is nominally 10 min or less.1.7 Improvements to this test method have been reported (1,2,3 ,4 ). Discussions of these improvements are not included in this test method.1.8 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific precautionary statements are given in Section 8.

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

ASTM C1128-01 核燃料循环材料分析用操作参考材料制备的标准指南

Certified reference materials (CRMs) prepared from nuclear materials are generally of high purity, possessing chemical stability or reproducible stoichiometry. Usually they are certified using the most unbiased and precise measurement methods available, often with more than one laboratory being involved in making certification measurements. CRMs are generally used on a national or international level, and they are at the top of the metrological hierarchy of reference materials. A graphical representation of a national nuclear measurement system is shown in Fig. 3. Working reference materials (WRMs) need to have quality characteristics that are similar to CRMs, although the rigor used to achieve those characteristics is not usually as stringent as for CRMs. Where possible, CRMs are often used to calibrate the methods used for establishing the concentration values (reference values) assigned to WRMs, thus providing traceability to CRMs as required by ISO 17025. A WRM is normally prepared for a specific application. Because of the importance of having highly reliable measurement data from nuclear materials, particularly for control and accountability purposes, CRMs are sometimes used for calibration when available. However, CRMs prepared from nuclear materials are not always available for specific applications. Thus, there may be a need for a laboratory to prepare WRMs from nuclear materials. Also, CRMs are often too expensive, or their supply is too limited for use in the quantities needed for long-term, routine use. When properly prepared, WRMs will serve equally well as CRMs for most applications, and using WRMs will preserve supplies of CRMs. Difficulties may be encountered in the preparation of RMs from nuclear materials because of the chemical and physical properties of the materials. Chemical instabilities, problems in ensuring stoichiometry, and radioactivity are factors involved, with all three factors being involved with some materials. Those preparing WRMs from nuclear materials must be aware of how these factors affect preparation, as well as being aware of the other criteria governing the preparation of reliable WRMs. FIG. 3 United States Nuclear Measurement System1.1 This guide covers the preparation and characterization of working reference materials (WRM) that are produced by a laboratory for its own use in the analysis of nuclear materials. Guidance is provided for establishing traceability of WRMs to certified reference materials by a defined characterization process. The guidance provided is generic; it is not specific for a given material.1.2 The information provided by this guide is found in the following sections:SectionPlanning6Preparation7Packaging and Storage8Characterization9Statistical Analysis10Documentation 111.3 The values stated in SI units are to be regarded as the 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 Guide for Preparation of Working Reference Materials for Use in the Analysis of Nuclear Fuel Cycle Materials

ASTM C1316-01 使用a252Cf的被动-主动中子计数法在碎片和废料中作核材料无损检定的标准试验方法

1.1 This test method covers the nondestructive assay of scrap and waste for uranium and plutonium content using a 252Cf shuffler. Shuffler measurements provide rapid results and can be applied to a variety of matrix materials in containers as large as 208-litre drums. Corrections are made for the effects of matrix material. This test method has been used to assay items containing uranium, plutonium, or both. Applications of this test method include measurements for safeguards, accountability, TRU, and U waste segregation, disposal, and process control purposes (1,2,3).1.1.1 This test method uses passive neutron coincidence counting to measure 238Pu, 240Pu, and 242Pu. It has been used to assay items with plutonium contents between 0.03 g and 1000 g. It could be used to measure other spontaneously fissioning isotopes. It specifically describes the approach used with shift register electronics; however, it can be adapted to other electronics. 1.1.2 This test method uses neutron irradiation with a moveable californium source and counting of the delayed neutrons from the induced fissions to measure 235U. It has been used to assay items with 235U contents between 0.1 g and 1000 g. It could be used to assay other fissionable isotopes.1.2 This test method requires knowledge of the relative isotopic composition to determine the mass of the different elements.1.3 This test method may give biased results for measurements of containers that include large quantities of hydrogen.1.4 The techniques described in this test method have been applied to materials other than scrap and waste. These other applications are not addressed in this test method.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. Specific precautionary statements are given in Section 8.

Standard Test Method for Nondestructive Assay of Nuclear Material in Scrap and Waste by Passive-Active Neutron Counting Using a 252 Cf Shuffler

ASTM C1502-01 测定二氧化铀和氧化钆中氯和氟的总含量的标准试验方法

The method is designed to show whether or not the tested materials meet the specifications as given in either Specification C 753, C 776, C 888 or C 922.1.1 This test method covers the determination of chlorine and fluorine in nuclear-grade uranium dioxide (UO2) powder and pellets, nuclear grade gadolinium oxide (Gd2O3) powder and gadolinium oxide-uranium oxide (Gd2O3-UO2) powder and pellets.1.2 With a 2 gram UO2 sample size the detection limit of the method is 4 956;g/g for chlorine and 2 956;g/g for fluorine. The maximum concentration determined with a 2 gram sample is 500 956;g/g for both chlorine and fluorine. The sample size used in this test method can vary from 1 to 10 grams resulting in a corresponding change in the detection limits and range.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.

Standard Test Method for Determination of Total Chlorine and Fluorine in Uranium Dioxide and Gadolinium Oxide

ASTM C1348-01 直接氢还原成核级二氧化铀用带低于5%的235U含量的混合氧化铀用标准规范

1.1 This specification covers blended uranium trioxide (UO3), U3O8, or mixtures of the two, powders that are intended for conversion into a sinterable uranium dioxide (UO2) powder by means of a direction reduction process. The UO2 powder product of the reduction process must meet the requirements of Specification C753 and be suitable for subsequent UO2 pellet fabrication by pressing and sintering methods. This specification applies to uranium oxides with a 235U enrichment less than 5 %. 1.2 This specification includes chemical, physical, and test method requirements for uranium oxide powders as they relate to the suitability of the powder for storage, transportation, and direct reduction of UO2 powder. This specification is applicable to uranium oxide powders for such use from any source. 1.3 The scope of this specification does not comprehensively cover all provisions for preventing criticality accidents, for health and safety, or for shipping. Observance of this specification does not relieve the user of the obligation to conform to all international, national, state, and local regulations for processing, shipping, or any other way of using uranium oxide powders (see 2.2 and 2.3). 1.4 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only. 1.5 The following safety hazards caveat pertains only to the test methods portion of the annexes in 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 Blended Uranium Oxides with a ²³⁵U Content of Less Than 5 % for Direct Hydrogen Reduction to Nuclear Grade Uranium Dioxide

EJ/T 1121-2000 铀矿样品加工和管理技术规范

本标准规定了铀矿样品加工和样品管理的技术要求。 本标准适用于放射性矿产地质分析测试实验室的样品加工和管理，适用于实验室的质量管 理。

Standard of manufacturing and managerial technique for the Uranium samples

DIN 25471:2000 带轻水反应堆的核电站燃料槽中核燃料元件在搬运和储存时燃耗应考虑的临界安全性

The document specifies requirements which guarantee an observation of criticality safety in handling and storage of nuclear fuel elements in fuel storage pools of nuclear power stations with lightwater reactors when during planning, setting up or during operation of corresponding handling and storage devices the burn up of the fuel elements is taken into account.#,,#

Criticality safety taking into account the burnup of fuel elements when handling and storing of nuclear fuel elements in fuel pools of nuclear power stations with lightwater reactors

NF M60-421:2000 核能.用溶剂色谱法对同位素和稀释度分析用的硝酸溶液中铀和钚的化学分离和提纯

Nuclear energy - Chemical separation and purification of uranium and plutonium in nitric acid solutions for isotopic and dilution analysis by solvent chromatography.

ASTM C1284-00(2005) 用α光谱测定法电镀锕元素的标准规程

The determination of actinide elements by alpha spectrometry measurement is an essential part of many environmental research and monitoring programs. Alpha spectrometry measurements identify and quantify the alpha-emitting actinide elements. A variety of separation methods will typically preceed the electrodeposition of a sample for alpha spectrometry measurements. In addition to the electrodeposition procedure presented in this practice, the scientific literature contains other procedures for actinide electrodeposition. Note 18212;An alternate method for mounting actinides for alpha spectrometry measurements by coprecipitation with neodymium fluoride is described in Test Methods C 1163.1.1 This practice covers the preparation of separated actinide fractions for alpha spectrometry measurement. It is applicable to any of the actinides that can be dissolved in dilute ammonium sulfate solution. Examples of applicable actinide fractions would be the final elution from an ion exchange separation or the final strip from a solvent extraction separation. 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 Practice for Electrodeposition of the Actinides for Alpha Spectrometry

ASTM C1462-00 大于15%小于20%的235U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 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 C1474-00(2006)e1 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

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 , 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......

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

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 E263-00 用铁的放射性测量快中子反应速率的标准试验方法

1.1 This test method describes procedures for measuring reaction rates by the activation reaction Fe (n,p) Mn. 1.2 This activation reaction is useful for measuring neutrons with energies above approximately 2.2 MeV and for irradiation times up to about 3 years (for longer irradiations, see Practice E261). 1.3 With suitable techniques, fission-neutron fluence rates above 10 cm [dot]s can be determined. However, in the presence of a high thermal-neutron fluence rate (for example, >2 X 10 cm [dot]s , Mn depletion should be investigated. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261. 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 Fast-Neutron Reaction Rates by Radioactivation of Iron

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 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 C1284-00 用α光谱测定法电镀锕元素的标准规程

1.1 This practice covers the preparation of separated actinide fractions for alpha spectrometry measurement. It is applicable to any of the actinides that can be dissolved in dilute ammonium sulfate solution. Examples of applicable actinide fractions would be the final elution from an ion exchange separation or the final strip from a solvent extraction separation. 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 Practice for Electrodeposition of the Actinides for Alpha Spectrometry

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