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

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 C1217-00(2012) 核材料及放射性材料处理用设备设计的标准指南

Equipment operability and long-term integrity are concerns that originate during the design and fabrication sequences. Such concerns can only be addressed or are most efficiently addressed during one or the other of these stages. Equipment operability and integrity can be compromised during handling and installation sequences. For this reason, the subject equipment should be handled and installed under closely controlled and supervised conditions. This guide is intended as a supplement to other standards, and to federal and state regulations, codes, and criteria applicable to the design of equipment intended for this use. This guide is intended to be generic and to apply to a wide range of equipment types and configurations. The term equipment is used herein in a generic sense. See 3.2.5 for the definition. This service imposes stringent requirements on the quality and the integrity of the equipment, as follows: Leak tightness is required. This implies containment of liquids at all times, and retention of vapors and gases by means of vessel design, or through means of engineered provisions or operational procedures, or both, that ensure the retention, collection, and treatment of vapors and off-gases when the vessel cannot be fabricated or operated with an air-tight vessel configuration. Radioactive materials must be contained. Equipment must be capable of withstanding rigorous chemical cleaning and decontamination procedures. Equipment must be designed and fabricated to remain dimensionally stable throughout its life cycle. Close fabrication tolerances are required to set nozzles and other datum points in known positions. Fabrication materials must be resistant to radiation damage, or materials subject to such damage must be shielded or placed so as to be readily replaceable. Smooth surface finishes are required. Irregularities that hide and retain radioactive particulates or other adherent contamination must be eliminated. Equipment must be capable of being operated virtually unattended, unseen, and trouble-free over long periods. It is assumed that the radiation hazards, combined with the need for confinement and containment, will necessitate a shielded enclosure cell equipped for some degree of remote handling and processing capability in the transuranic materials handling, processing, or machining operations (see 1.2.2). Equipment intended for use in the processing and incorporation of radioactive wastes in host composites or matrices may operate at high temperatures and pressures and may require engineered provisions for the removal of large heat loads under normal and emergency conditions. The chemical corrosion and erosion conditions encountered in these processes tend to be extremely severe, placing emphasis on design for containment integrity. Maintenance records from the plant or from a plant having a similar processing mission may be available for reference. If available and accessible, these records may offer valuable insight with regard to the causes, frequency, and type of failure experienced for the type and class of equipment being designed and engineered. The constraints cited herein are intended to help the engineer establish conditions aimed toward the following: Enhancing radioactive materials containment integrity, Minimizing the loss of in-process materials or the spread of hazardous radioactive contaminants, Minimizing equipment blemishes or faults that promote the adherence or retention of radiation sources, Facilitating the ease and safety of decontamination and maintenance sequences, and Reducing the failure frequency rate for all types and classes of equipment used in this service. Exclusions: In general, this guide is not intended......

Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials

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 C1062-00 核燃料溶解装置的设计、制造和安装标准指南

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 Exclusions1.2.1 Operations-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 Procedures-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 Processes-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 Pretreatment-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 judged essential to an efficient nuclear fuels dissolution step.1.2.5 Fundamentals- This guide does not address specific chemical, physical or mechanical technology, fluid mechanics, stress analysis or other engineering fundamentals that are also applied in the creation of a safe design for nuclear fuel dissolution facilities.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 Guide for Design, Fabrication, and Installation of Nuclear Fuel Dissolution Facilities

ASTM E2089-00 太空用材料的地面实验室原子氧交互评价的标准规程

1.1 The intent of these practices is to define atomic oxygen exposure procedures that are intended to minimize variability in results within any specific atomic oxygen exposure facility as well as contribute to the understanding of the differences in the response of materials when tested in different facilities.1.2 These practices are not intended to specify any particular type of atomic oxygen exposure facility but simply specify procedures that can be applied to a wide variety of facilities.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 The values stated in SI units are to be regarded as the standard.

Standard Practices for Ground Laboratory Atomic Oxygen Interaction Evaluation of Materials for Space Applications

ASTM C1462-00(2013) 大于15%且小于20%浓缩铀金属U-235的标准规格

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 208201;% (and greater than 158201;%) 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, C996 regarding references).

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

ASTM C1267-00 使用铁II在磷酸中还原后在有钒的情况下滴铬(VI)的铀标准试验方法

1.1 This test method, commonly referred to as the Modified Davies and Gray technique, covers the titration of uranium in product, fuel, and scrap materials after the material is dissolved. The test method is versatile and has been ruggedness tested. With appropriate sample preparation, this test method can give precise and unbiased uranium assays over a wide variety of material types (1, 2). Details of the titration procedure in the presence of plutonium with appropriate modifications are given in Test Method C1204. 1.2 Uranium levels titrated are usually 20 to 50 mg, but up to 200 mg uranium can be titrated using the reagent volumes stated in this test method. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. For specific safeguard and safety precaution statements, see Section 4.

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

ASTM E2089-00(2006) 太空用材料的地面实验室原子氧交互评价的标准规程

These practices enable the following information to be available: 3.1.1 Material atomic oxygen erosion characteristics. 3.1.2 An atomic oxygen erosion comparison of four well-characterized polymers. The resulting data are useful to: 3.2.1 Compare the atomic oxygen durability of spacecraft materials exposed to the low Earth orbital environment. 3.2.2 Compare the atomic oxygen erosion behavior between various ground laboratory facilities. 3.2.3 Compare the atomic oxygen erosion behavior between ground laboratory facilities and in-space exposure. 3.2.4 Screen materials being considered for low Earth orbital spacecraft application. However, caution should be exercised in attempting to predict in-space behavior based on ground laboratory testing because of differences in exposure environment and synergistic effects.1.1 The intent of these practices is to define atomic oxygen exposure procedures that are intended to minimize variability in results within any specific atomic oxygen exposure facility as well as contribute to the understanding of the differences in the response of materials when tested in different facilities.1.2 These practices are not intended to specify any particular type of atomic oxygen exposure facility but simply specify procedures that can be applied to a wide variety of facilities.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 The values stated in SI units are to be regarded as the standard.

Standard Practices for Ground Laboratory Atomic Oxygen Interaction Evaluation of Materials for Space Applications

ASTM C1217-00(2006) 核材料及放射性材料处理用设备设计的标准导则

1.1 Intent1.1.1 This guide covers equipment used in shielded cell or canyon facilities for the processing of nuclear and radioactive materials. It is the intent of this guide to set down the conditions and practices that have been found necessary to ensure against or to minimize the failures and outages of equipment used under the subject circumstances.1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, and installation of equipment capable of meeting the stringent demands of operating, dependably and safely, in a nuclear processing environment that operators can neither see nor reach directly.1.1.3 This guide sets forth generalized criteria and guidelines for the design, fabrication, and installation of equipment used in this service. This service includes the processing of radioactive wastes. Equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. In the type of shielded cell or canyon facility of interest to users of this guide, either the background radiation level remains high at all times or it is impractical to remove the process sources of radiation to facilitate in situ repairs or carry out maintenance procedures on equipment. The equipment is operated remotely, either with or without visual access to the equipment.1.2 Applicability1.2.1 This guide is intended to be applicable to equipment used under one or more of the following conditions:The materials handled or processed constitute a significant radiation hazard to man or to the environment.The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded.The material handled or processed must be retained, contained, and confined within known bounds for reasons of accountability or to minimize the spread of radioactive contamination.The materials handled or processed must be kept and maintained within one or more of the following conditions:In a specific geometric array or configuration, andWithin a range of conditions that have been determined to be a critically safe set of conditions for that piece of equipment, that is, 1) in a given and specified operational position where adjacent nuclear criticality interaction conditions are known and unchanging, 2) for a given and specified set or range of operating conditions, and 3) for a given and specified process.The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without intervening shielded viewing windows, periscopes, or a television monitoring system.1.2.2 This guide is intended to be applicable to the design of equipment for the processing of materials containing uranium and transuranium elements in any physical form under the following conditions:Such materials constitute an unacceptable radiation hazard to the operators and maintenance personnel,The need exists for the confinement of the in-process material, of dusts and particulates, or of vapors and gases arising or resulting from the handling and processing of such materials, andAny of the conditions cited in apply.1.2.3 This guide is intended to apply to the design, fabrication, and installation of ancillary and support services equipment under the following conditions:Such equipment is installed in shielded cell or canyon environments, orSuch equipment is an integral part of an in-cell processing equipment configuration, or an auxiliary component or system thereof, even though an equipment item or system may not directly hold or contain nuclear or radioactive materials under normal processing conditions.

Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials

ASTM E2089-00(2014) 用于空间应用的地面实验室原子氧气交互评估的标准实施规程

3.1 These practices enable the following information to be available: 3.1.1 Material atomic oxygen erosion characteristics. 3.1.2 An atomic oxygen erosion comparison of four well-characterized polymers. 3.2 The resulting data are useful to: 3.2.1 Compare the atomic oxygen durability of spacecraft materials exposed to the low Earth orbital environment. 3.2.2 Compare the atomic oxygen erosion behavior between various ground laboratory facilities. 3.2.3 Compare the atomic oxygen erosion behavior between ground laboratory facilities and in-space exposure. 3.2.4 Screen materials being considered for low Earth orbital spacecraft application. However, caution should be exercised in attempting to predict in-space behavior based on ground laboratory testing because of differences in exposure environment and synergistic effects. 1.1 The intent of these practices is to define atomic oxygen exposure procedures that are intended to minimize variability in results within any specific atomic oxygen exposure facility as well as contribute to the understanding of the differences in the response of materials when tested in different facilities. 1.2 These practices are not intended to specify any particular type of atomic oxygen exposure facility but simply specify procedures that can be applied to a wide variety of facilities. 1.3 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 Practices for Ground Laboratory Atomic Oxygen Interaction Evaluation of Materials for Space Applications

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 C776-00 烧结的二氧化铀丸标准规范

1.1 This specification is for finished sintered uranium dioxide pellets. It applies to uranium dioxide pellets containing uranium of any 234U concentration for use in nuclear reactors. 1.2 This specification does not include (a) provisions for preventing criticality accidents or (b) 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 federal, state, and local regulations pertaining to possessing, shipping, processing, or using source or special nuclear material. Examples of U.S. Government documents are Code of Federal Regulations (Latest Edition), Title 10, Part 50, Title 10, Part 71, and Title 49, Part 173. 1.3 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 or regulatory limitations prior to use.

Standard Specification for Sintered Uranium Dioxide Pellets

ASTM C1456-00 用X射线荧光法测定氧化钆和氧化铀颗粒中铀和/或钆的标准试验方法

This guide is applicable to samples containing 2 to 10 % gadolinium oxide and 90 to 98 % uranium oxide on the “as received” basis. The method may be used to determine concentration of either uranium, gadolinium, or both. Either wavelength-dispersive or energy-dispersive x-ray fluorescence systems may be used provided the software accompanying the system is able to accommodate the use of internal standards.1.1 This standard describes the steps necessary for the preparation and analysis by X-ray fluorescence (XRF) of gadolinium and/or uranium in gadolinium oxide-uranium oxide pellets or powders.1.2 This method requires the use of appropriate internal standard(s). Care must be taken to ascertain that samples analyzed by this method do not contain the internal standard element(s) or that this contamination has been corrected for mathematically whenever present. Such corrections are not addressed in this 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. Specific precautions are given in Section and various notes throughout the method.

Standard Test Method for the Determination of Uranium or Gadolinium, or Both, in Gadolinium Oxide-Uranium Oxide Pellets or by X-Ray Fluorescence (XRF)

ASTM C1062-00(2014) 核燃料溶解设施的设计, 制造和安装标准指南

4.1 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 Operations—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 Procedures—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 Processes—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 Pretreatment—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 disasse......

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

ASTM C1217-00 核材料及放射性材料处理用设备设计的标准导则

1.1 Intent: 1.1.1 This guide covers equipment used in shielded cell or canyon facilities for the processing of nuclear and radioactive materials. It is the intent of this guide to set down the conditions and practices that have been found necessary to ensure against or to minimize the failures and outages of equipment used under the subject circumstances. 1.1.2 It is intended that this guide record the principles and caveats that experience has shown to be essential to the design, fabrication, and installation of equipment capable of meeting the stringent demands of operating, dependably and safely, in a nuclear processing environment that operators can neither see nor reach directly. 1.1.3 This guide sets forth generalized criteria and guidelines for the design, fabrication, and installation of equipment used in this service. This service includes the processing of radioactive wastes. Equipment is placed behind radiation shield walls and cannot be directly accessed by the operators or by maintenance personnel because of the radiation exposure hazards. In the type of shielded cell or canyon facility of interest to users of this guide, either the background radiation level remains high at all times or it is impractical to remove the process sources of radiation to facilitate in situ repairs or carry out maintenance procedures on equipment. The equipment is operated remotely, either with or without visual access to the equipment. 1.2 Applicability: 1.2.1 This guide is intended to be applicable to equipment used under one or more of the following conditions: 1.2.1.1 The materials handled or processed constitute a significant radiation hazard to man or to the environment. 1.2.1.2 The equipment will generally be used over a long-term life cycle (for example, in excess of two years), but equipment intended for use over a shorter life cycle is not excluded. 1.2.1.3 The material handled or processed must be retained, contained, and confined within known bounds for reasons of accountability or to minimize the spread of radioactive contamination. 1.2.1.4 The materials handled or processed must be kept and maintained within one or more of the following conditions: (1) In a specific geometric array or configuration, and (2) Within a range of conditions that have been determined to be a critically safe set of conditions for that piece of equipment, that is, 1) in a given and specified operational position where adjacent nuclear criticality interaction conditions are known and unchanging, 2) for a given and specified set or range of operating conditions, and 3) for a given and specified process. 1.2.1.5 The equipment can neither be accessed directly for purposes of operation or maintenance, nor can the equipment be viewed directly, for example, without intervening shielded viewing windows, periscopes, or a television monitoring system. 1.2.2 This guide is intended to be applicable to the design of equipment for the processing of materials containing uranium and trans-uranium elements in any physical form under the following conditions: 1.2.2.1 Such materials constitute an unacceptable radiation hazard to the operators and maintenance personnel, 1.2.2.2 The need exists for the confinement of the in-process material, of dusts and particulates, or of vapors and gases arising or resulting from the handling and processing of such materials, and 1.2.2.3 Any of the conditions cited in 1.2.1 apply. 1.2.3 This guide is intended to apply to the design, fabrication, and installation of ancillary and support services equipment under the following conditions: 1.2.3.1 Such equipment is installed in shielded cell or canyon environments, or 1.2.3.2 Such equipment is an integral part of an in-cell processing equipment configuration, or an auxiliary component or system thereof, even though an equipment ite......

Standard Guide for Design of Equipment for Processing Nuclear and Radioactive Materials

ASTM C1411-99 同位素分析前铀和钚离子交换分离的标准实施规程

1.1 This practice is for the ion exchange separation of uranium and plutonium from each other and from other impurities for subsequent isotopic analysis by thermal ionization mass spectrometry. Plutonium-238 and uranium-238, and plutonium-241 and americium-241, will appear as the same mass peak and must be chemically separated prior to analysis. Only high purity solutions can be analyzed reliably using thermal ionization mass spectrometry. 1.2 This standard may involve hazardous material, 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.

Standard Practice for the Ion Exchange Separation of Uranium and Plutonium Prior to Isotopic Analysis

ASTM C1429-99(2014) 采用双标准多收集器气体质谱仪进行六氟化铀同位素分析的标准试验方法

5.1 Uranium hexafluoride used to produce nuclear-reactor fuel must meet certain criteria for its isotopic composition. This test method may be used to help determine if sample materials meet the criteria described in Specifications C787 and C996. 1.1 This test method covers a quantitative test method applicable to determining the mass percent of uranium isotopes in uranium hexafluoride (UF6) samples. This method as described is for concentrations of 235U between 0.1 and 10 mass8201;%, and 234U and 236U between 0.0001 and 0.1 mass8201;%. 1.2 This test method is for laboratory analysis by a gas mass spectrometer with a multi-collector. 1.3 This standard complements Test Methods C761, the double-standard method for gas mass spectrometers using a single collector, by providing a method for spectrometers using a multi-collector. 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 Isotopic Analysis of Uranium Hexafluoride by Double-Standard Multi-Collector Gas Mass Spectrometer

ASTM C1428-99(2005) 用单支架气体源多收集器质谱仪法进行六氟化铀的同位素分析的标准试验方法

Uranium hexafluoride is a basic material used to produce nuclear reactor fuel. To be suitable for this purpose, the material must meet criteria for isotopic composition. This test method is designed to determine whether the material meets the requirements described in Specifications C 787 and C 996.1.1 This test method is applicable to the isotopic analysis of uranium hexafluoride (UF6) with 235U concentrations less than or equal to 5 % and 234U, 236U concentrations of 0.001 to 0.1 %. 1.2 This test method may be applicable to the analysis of the entire range of 235U isotopic compositions providing that adequate standards are available. 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 Isotopic Analysis of Uranium Hexafluoride by Single-Standard Gas Source Multiple Collector Mass Spectrometer Method

ASTM C1431-99 维护容器清理用铝基废核燃料腐蚀检验的标准导则

1.1 This guide covers corrosion testing of aluminum-based spent nuclear fuel in support of geologic repository disposal (per the requirements in 10 CFR 60 and 40 CFR 191). The testing described in this document is designed to provide data for analysis of the chemical stability and radionuclide release behavior of aluminum-based waste forms produced from aluminum-based spent nuclear fuels. The data and analyses from the corrosion testing will support the technical basis for inclusion of aluminum-based spent nuclear fuels in the repository source term. Interim storage and transportation of the spent fuel will precede geologic disposal; therefore, reference is also made to the requirements for interim storage (per 10 CFR 72) and transportation (per 10 CFR 71). The analyses that will be based on the data developed are also necessary to support the safety analyses reports (SARs) and performance assessments (PAs) for disposal systems. 1.2 Spent nuclear fuel that is not reprocessed must be safely managed prior to transportation to, and disposal in, a geologic repository. Placement is an interim storage facility may include direct placement of the irradiated fuel or treatment of the fuel prior to placement, or both. The aluminum-based waste forms may be required to be ready for geologic disposal, or road ready, prior to placement in extended interim storage. Interim storage facilities, in the United States, handle fuel from civilian commercial power reactors, defense nuclear materials production reactors, and research reactors. The research reactors include both foreign and domestic reactors. The aluminum-based fuels in the spent fuel inventory in the United States are primarily from defense reactors and from foreign and domestic research reactors. The aluminum-based spent fuel inventory includes several different fuel forms and levels of 235U enrichment. Highly enriched fuels (235U enrichment leves 062 20%) are part of this inventory. 1.3 Knowledge of the corrosion behavior of aluminum-based spent nuclear fuels is required to ensure safety and to support licensing or other approval activities, or both, necessary for disposal in a geologic repository. The response fo the aluminum-based spent nuclear fuel waste form(s) to disposal environments must be established for configuration-safety analyses, criticality analyses, PAs, and other analyses required to assess storage, treatment, transportation, and disposal of spent nuclear fuels. This is particularly important for the highly enriched, aluminum-based spent nuclear fuels. The test protocols described in this guide are designed to establish material response under the repository relevant conditions. 1.4 The majority of the aluminum-based spent nuclear fuels are aluminum clad, aluminum-uranium alloys. The aluminum-uranium alloy typically consists of uranium aluminide particles dispersed in an aluminum matrix. Other aluminum-based fuels include dispersions of uranium oxide, uranium silicide, or uranium carbide particles in an aluminum matrix. These particles, including the aluminides, are generally cathodic to the aluminum matrix. Selective leaching of the aluminum in the exposure environment may provide a mechanism for redistribution and relocation of the uranium-rich particles. Particle redistribution tendencies will depend on the nature of the aluminum corrosion processes and the size, shape, distribution and relative reactivity of the uranium-rich particles. Interpretation of test data will require an understanding of the material behavior. This understanding will enable evaluation of the design and configuration of the waste package to ensure that unfilled regions in the waste package do not provide sites for the relocation of the uranium-rich particles into nuclear critical configurations. Test samples must be evaluated, prior to testing, to ensure that the size and shape of the uranium-rich particles in the test samples are......

Standard Guide for Corrosion Testing of Aluminum-Based Spent Nuclear Fuel in Support of Repository Disposal

ASTM C1254-99 用X射线荧光法测定无机酸中铀的标准试验法

1.1 This test method covers the steps necessary for the preparation and analysis by X-ray fluorescence (XRF) of mineral acid solutions containing uranium. 1.2 This test method is valid for those solutions containing 2 to 20 g uranium/L as presented to the spectrometer. Higher concentrations may be covered by appropriate dilutions. 1.3 This test method requires the use of an appropriate internal standard. Care must be taken to ascertain that samples analyzed by this test method do not contain the internal standard element or that this contamination has been corrected for mathematically whenever present. Such corrections are not addressed in this test method. Care must also be taken that the choice of internal standard and sample medium are compatible; that is, do not use yttrium with solutions containing HF or strontium with those having H 2 SO 4 . Alternatively a scatter line may be used as internal standard. 1.4 The values stated in SI units are to be regarded as the standard. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. Specific precautionary statements are given in Section 8 and Note 1.

Standard Test Method for Determination of Uranium in Mineral Acids by X-Ray Fluorescence