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

ASTM C1458-09 采用量热法对钚, 氚和241Am进行无损测量的标准试验方法

This test method is the most accurate NDA technique for the assay of many physical forms of Pu. Isotopic measurements by gamma-ray spectroscopy or destructive analysis techniques are part of the test method when it is applied to the assay of Pu. Calorimetry has been applied to a wide variety of Pu-bearing solids including metals, alloys, oxides, fluorides, mixed Pu-U oxides, mixed oxide fuel pins, waste, and scrap, for example, ash, ash heels, salts, crucibles, and graphite scarfings) (2,3). This test method has been routinely used at U.S. and European facilities for Pu process measurements and nuclear material accountability for the last 40 years (2-9). Pu-bearing materials have been measured in calorimeter containers ranging in size from about 0.025 m to about 0.60 m in diameter and from about 0.076 m to about 0.9 m in height. Gamma-ray spectroscopy typically is used to determine the Pu-relative isotopic composition and 241Am to Pu ratio (see Test Method C 1030). Isotopic information from mass spectrometry and alpha counting measurements may be used (see Test Method C 697). This test method is the most accurate NDA method for the measurement of tritium. For many physical forms of tritium compounds calorimetry is the only practical measurement technique available. Physical standards representative of the materials being assayed are not required for the test method. This test method is largely independent of the elemental distribution of the nuclear materials in the matrix. The accuracy of the method can be degraded for materials with inhomogeneous isotopic composition. The thermal power measurement is traceable to national measurement systems through electrical standards used to directly calibrate the calorimeters or to calibrate secondary 238Pu heat standards. Heat-flow calorimetry has been used to prepare secondary standards for neutron and gamma-ray assay systems (7-12). Calorimetry measurement times are typically longer than other NDA techniques. Four parameters of the item and the item packaging affect measurement time. These four parameters are density, mass, thermal conductivity, and change in temperature. The measurement well of passive calorimeters will also affect measurement time because it too will need to come to the new equilibrium temperature. Calorimeters operated in servo mode maintain a constant measurement well temperature and have no effect on measurement time. Calorimeter measurement times range from 20 minutes (13) for smaller, temperature-conditioned, containers up to 24 h for larger containers and items with long thermal-time constants. Measurement times may be reduced by using equilibrium prediction techniques, by temperature preconditioning of the item to be measured, or operating the calorimeter using the servo-control technique.1.1 This test method describes the nondestructive assay (NDA) of plutonium, tritium, and 241Am using heat flow calorimetry. For plutonium the typical range of applicability corresponds to ~1 g to ~2000 g quantities while for tritium the typical range extends from ~0.001 g to~ 10 g. This test method can be applied to materials in a wide range of container sizes up to 50 L. It has been used routinely to assay items whose thermal power ranges from 0.001 W to 135 W. 1.2 This test method requires knowledge of the relative abundances of the plutonium isotopes and the 241Am/Pu mass ratio to determine the total plutonium mass. 1.3 This test method provides a direct measure of tritium content. 1.4 This test method provides a measure of 241Am either as a single isotope or mixed with plutonium. 1.......

Standard Test Method for Nondestructive Assay of Plutonium, Tritium and ²⁴¹Am by Calorimetric Assay

ASTM C750-09(2014) 核纯级碳化硼粉末的标准规格

1.1 This specification defines the chemical and physical requirements for boron carbide powder intended for a variety of nuclear applications. Because each application has a different need for impurity and boron requirements, three different chemical compositions of powder are specified. In using this specification, it is necessary to dictate which type of powder is intended to be used. In general, the intended applications for the various powder types are as follows: 1.1.1 Type 1—For use as particulate material in nuclear reactor core applications. 1.1.2 Type 2—Powder that will be further processed into a fabricated shape for use in a nuclear reactor core or used in non-core applications when the powder directly or indirectly may cause adverse effects on structural components, such as halide stress corrosion of stainless steel. 1.1.3 Type 3—Powder that will be used for non-core applications or special in-core applications. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

Standard Specification for Nuclear-Grade Boron Carbide Powder

ASTM C1493-09 用微分衰减系统钝化-活化中子计数法对废物中的核物质进行无损化验的标准试验方法

This test method is useful for quantifying fissile (for example, 233U, 235U, 239Pu and 241Pu) and spontaneously-fissioning nuclei (for example, 238Pu, 240Pu, 242Pu, 244Cm, 248Cm, and 252Cf) in waste and scrap drums. Total elemental mass of the radioactive materials can be calculated if the relative abundances of each radionuclide are known. Typically, this test method is used to measure one fissile isotope (for example, 235U or 239Pu). This test method can be used to segregate low level and transuranic waste at the 100 nCi/g concentration level currently required to meet the DOE Waste Isolation Pilot Plant (WIPP) waste acceptance criterion (5, 8, 9). This test method can be used for waste characterization to demonstrate compliance with the radioactivity levels specified in waste, disposal, and environmental regulations (See NRC regulatory guides, DOE Order 435.1, 10 CFR Part 71, 40 CFR Part 191, and DOE /WIPP-069). In the active mode, the DDT system can measure the 235U content in the range from <0.02 to >100 g and the 239Pu content, nominally between <0.01 and >20 g. In the passive mode, the DDT system is capable of assaying spontaneously-fissioning nuclei, over a nominal range from 0.05 to 15 g of 240Pu, or equivalent (5, 10, 11, 12, 13). This test method should be used in conjunction with a waste management plan that segregates the contents of assay items into material categories according to some or all of the following criteria: bulk density of the waste, chemical forms of the plutonium or uranium and matrix, (α, n) neutron intensity, hydrogen (moderator) and absorber content, thickness of fissile mass(es), and the assay item container size and composition. Each matrix may require a different set of calibration standards and may have different mass calibration limits. The effect on the quality of the assay (that is, minimizing precision and bias) can significantly depend on the degree of adherence to this waste management plan. The bias of the measurement results is related to the fill height, the homogeneity and composition of the matrix, the quantity and distribution of the nuclear material, and the item size. The precision of the measurement results is related to the quantity of the nuclear material, the background, and the count time of the measurement. For both matrix-specific and wide-range calibrations, this test method assumes the calibration material matches the items to be measured with respect to homogeneity and composition of the matrix, the neutron moderator and absorber content, and the quantity, distribution, and form of nuclear material, to the extent they affect the measurement. The algorithms for this test method assume homogeneity. Heterogeneity in the distribution of nuclear material, neutron moderators, and neutron absorbers has the potential to cause biased results (14). This test method assumes that the distribution of the contributing radioisotopes is uniform throughout the container and that lumps of nuclear material are not present. Reliable results from the application of this test method require waste to be packaged so the conditions of Section 5.5 can be met. In some cases, site-specific requirements will dictate the packaging requirements with possible detrimental effects to the measurement results. Both the active mode and the passive mode provide assay values for plutonium. During the calibration process, the operator should determine the applicable mass ranges for both modes of operation.1.1 This test method covers a sy......

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 C1411-08 同位素分析前铀和钚离子交换分离的标准实施规程

Uranium and plutonium are used in nuclear reactor fuel and must be analyzed to insure that they meet certain criteria for isotopic composition as described in Specification C 833 and Specification C 1008. This standard practice is used to chemically separate the same mass peak interferences from uranium and plutonium and from other impurities prior to isotopic abundance determination by thermal ionization mass spectrometry. In those facilities where perchloric acid use is tolerated, the separation in Test Method C 698 may be used prior to isotopic abundance determination. Uranium and plutonium concentrations as well as isotopic abundances using thermal ionization mass spectrometry can be determined using this separation and following Test Method C 1625.1.1 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 consult and 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 D7219-08 同位素和接近同位素核裂变物质标准规范

1.1 This specification covers the classification, processing, and properties of nuclear grade graphite billets with dimensions sufficient to meet the designer’s requirements for fuel elements, moderator or reflector blocks, in a high temperature gas cooled reactor. The graphite classes specified here would be suitable for reactor core applications where neutron irradiation induced dimensional changes are a significant design consideration. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. (See IEEE/ASTM SI 10.) 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 Specification for Isotropic and Near-isotropic Nuclear Graphites

ASTM C1477-08(2014) 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

5.1 The test method is capable of measuring uranium isotopic abundances of8201;234U,8201;235U,8201; 236U and8201;238U as required by Specifications C787 and C996. 1.1 This test method covers the isotopic abundance analysis of8201;234U,8201;235U,8201;236 U and8201;238U in samples of hydrolysed uranium hexafluoride (UF6) by inductively coupled plasma source, multicollector, mass spectrometry (ICP-MC-MS). The method applies to material with8201; 235U abundance in the range of 0.2 to 68201;% mass. This test method is also described in ASTM STP 1344. 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.

Standard Test Method for Isotopic Abundance Analysis of Uranium Hexafluoride and Uranyl Nitrate Solutions by Multi-Collector, Inductively Coupled Plasma-Mass Spectrometry

ASTM E704-08 用铀238的放射活化测定反应速率的测试方法

Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors. 238U is available as metal foil, wire, or oxide powder (see Guide E 844). It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the 238U and its fission products. One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2. 137Cs-137mBa is chosen frequently for long irradiations. Radioactive products 134Cs and 136Cs may be present, which can interfere with the counting of the 0.662 MeV 137Cs-137, Ba gamma rays (see Test Methods E 320). 140Ba-140La is chosen frequently for short irradiations (see Test Method E 393). 95Zr can be counted directly, following chemical separation, or with its daughter 95Nb using a high-resolution gamma detector system. 144Ce is a high-yield fission product applicable to 2- to 3-year irradiations. It is necessary to surround the 238U monitor with a thermal neutron absorber to minimize fission product production from a quantity of 235U in the 238U target and from 239Pu from (n,γ) reactions in the 238U material. Assay of the 239Pu concentration when a significant contribution is expected. Fission product production in a light-water reactor by neutron activation product 239Pu has been calculated to be insignificant (<2 %), compared to that from 238U(n,f), for an irradiation period of 12 years at a fast-neutron (E > 1 MeV) fluence rate of 1 × 1011 cm−2 · s−1 provided the 238U is shielded from thermal neutrons (see Fig. 2 of Guide E 844). Fission product production from photonuclear reactions, that is, (γ,f) reactions, while negligible near-power and research-reactor cores, can be large for deep-water penetrations (1). 1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 238U(n,f)F.P. 1.2 The reaction is useful for measuring neutrons with energies from approximately 1.5 to 7 MeV and for irradiation times up to 30 to 40 years. 1.3 Equivalent fission neutron fluence rates as defined in Practice E 261 can be determined. 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 unites of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238

ASTM C1689-08a(2014) 六氟化铀子二次采样的标准操作规程

5.1 Uranium hexafluoride is normally produced and handled in large (typically 1- to 14-ton) quantities and must, therefore be characterized by reference to representative samples. The samples are used to determine compliance with the applicable commercial Specifications C996 and C787 by means of the appropriate test method (for example, Test Method C761 and references therein). The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designated equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice indicates appropriate principles, equipment and procedures currently in use for subsampling of liquid UF6. It is used by UF6 converters, enrichers and fuel fabricators to review the effectiveness of existing procedures or to design equipment and procedures for future use. Other subsampling procedures such as UF6 vapor sampling are not directly representative of the chemical quality of liquid UF6. 5.2 It is emphasized that this test guide is not meant to address conventional or nuclear criticality safety issues. 1.1 This practice is applicable to subsampling uranium hexafluoride (UF6), using heat liquefaction techniques, from bulk containers, obtained in conformance with C1052, into smaller sample containers, which are required for laboratory analyses. 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 It is assumed that the liquid UF6 being sampled comprises a single quality and quantity of material. This practice does not address any special additional arrangement that might be required for taking proportional or composite samples. 1.4 The number of samples to be taken, their nominal sample weight, and their disposition shall be agreed upon between the parties. 1.5 The scope of this practice does not include provisions for preventing criticality incidents. 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 Practice for Subsampling of Uranium Hexafluoride

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 C1703-08(2013) 气态六氟化铀采样的标准实施规程

5.1 Uranium hexafluoride is normally produced and handled in large (typically 1 to 14-ton) quantities and must, therefore, be characterized by reference to representative samples (see ISO/DIS 7195). The samples are used to determine compliance with the applicable commercial specifications C996 and C787. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designed equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice can be used by UF6 converters, enrichers, and fuel fabricators to review the effectiveness of existing procedures or as a guide to the design of equipment and procedures for future use. 5.2 The intention of this practice is to avoid liquid UF6 sampling once the cylinder has been filled. For safety reasons, manipulation of large quantities of liquid UF6 should be avoided when possible. 5.3 It is emphasized that this practice is not meant to address conventional or nuclear criticality safety issues. 1.1 This practice covers methods for withdrawing representative sample(s) of uranium hexafluoride (UF6) during a transfer occurring in the gas phase. Such transfer in the gas phase can take place from a mother cylinder, for example in an autoclave to a receiving cylinder. It can also occur during the filling in the gas phase of a cylinder during a continuous production process, for example centrifuge enrichment facility or the distillation column in a conversion facility. Such sample(s) may be used for determining compliance with the applicable commercial specification, for example Specification C996 or Specification C787. 1.2 Since UF6 sampling is taken during the filling process, this practice does not address any special additional arrangements that may be agreed upon between the buyer and the seller when the sampled bulk material is being added to residues already present in a container (???heels recycle???). Such arrangements will be based on QA procedures such as traceability of cylinder origin (to prevent for example contamination with irradiated material). 1.3 If the receiving cylinder is purged after filling and sampling, special verifications must be performed by the user to verify the representativity of the sample(s). It is then expected that the results found on volatile impurities with gas phase sampling may be conservative. 1.4 This practice is only applicable when the transfer occurs in the gas phase. When the transfer is performed in the liquid phase, Practice

Standard Practice for Sampling of Gaseous Uranium Hexafluoride

ASTM C922-08 烧结氧化钆及二氧化铀颗粒的标准规范

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

Standard Specification for Sintered Gadolinium Oxide-Uranium Dioxide Pellets

ASTM E705-08 用镎237的放射活化测定反应速率的标准试验方法

Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors. 237Np is available as metal foil, wire, or oxide powder. For further information, see Guide E 844. It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the 237Np and its fission products. One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2. 137Cs-137mBa is chosen frequently for long irradiations. Radioactive products 134Cs and 136Cs may be present, which can interfere with the counting of the 0.662 MeV 137Cs-137mBa gamma ray (see Test Methods E 320). 140Ba-140La is chosen frequently for short irradiations (see Test Method E 393). 95Zr can be counted directly, following chemical separation, or with its daughter 95Nb, using a high-resolution gamma detector system. 144Ce is a high-yield fission product applicable to 2- to 3-year irradiations. It is necessary to surround the 237Np monitor with a thermal neutron absorber to minimize fission product production from trace quantities of fissionable nuclides in the 237Np target and from 238Np and 238Pu from (n,γ) reactions in the 237Np material. Assay of 238Pu and 239Pu concentration is recommended when a significant contribution is expected. Fission product production in a light-water reactor by neutron activation products 238Np and 238Pu has been calculated to be insignificant (1.2 %), compared to that from 237Np(n,f), for an irradiation period of 12 years at a fast neutron (E > 1 MeV) fluence rate of 1 × 1011 cm−2·s−1, provided the 237Np is shielded from thermal neutrons (see Fig. 2 of Guide E 844). Fission product production from photonuclear reactions, that is, (γ,f) reactions, while negligible near-power and researchreactor cores, can be large for deep-water penetrations (1). Good agreement between neutron fluence measured by 237Np fission and the 54Fe(n,p)54Mn reaction has been demonstrated (2). The reaction 237Np(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors. The 237Np fission neutron spectrum-averaged cross section in several benchmark neutron fields are given in Table 3 of Practice E 261. Sources for the latest recommended cross sections are given in Guide E 1018. In the case of the 237Np(n,f)F.P. reaction, the recommended cross section source is the ENDF/B-VI cross section (MAT = 9346) revision 1 (3). Fig. 1 shows a plot of the recommended cross section versus neutron energy for the fast-neutron reaction 237Np(n,f)F.P. Note 18212;The data are taken from the Evaluated Nuclear Data file, ENDF/B-VI, rather than the later ENDF/B-VII. This is in accordance with Guide E 1018 Guide for Application of ASTM Evaluated Cross Section Data File, 6.1. since the later ENDF/B-VII data files do not include covariance information. For more details s......

Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237

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 E262-08 用放射性技术测定热中子反应和注量率的标准试验方法

This test method can be extended to use any material that has the necessary nuclear and activation properties that suit the experimenter''s particular situation. No attempt has been made to fully describe the myriad problems of counting techniques, neutron-fluence depression, and thick-foil self-shielding. It is assumed that the experimenter will refer to existing literature on these subjects. This test method does offer a referee technique (the standard gold foil irradiation at National Institute of Standards and Technology (NIST) to aid the experimenter when he is in doubt of his ability to perform the radiometric technique with sufficient accuracy. The standard comparison technique uses a set of foils that are as nearly identical as possible in shape and mass. The foils are fabricated from any material that activates by an (n, γ) reaction, preferably having a cross section approximately inversely proportional to neutron speed in the thermal energy range. Some of the foils are irradiated in a known neutron field (at NIST) or other standards laboratory). The foils are counted in a fixed geometry on a stable radiation-detecting instrument. The neutron induced reaction rate of the foils is computed from the counting data, and the ratio of the known neutron fluence rate to the computed reaction rate is determined. For any given foil, neutron energy spectrum, and counting set-up, this ratio is a constant. Other foils from the identical set can now be exposed to an unknown neutron field. The magnitude of the fluence rate in the unknown field can be obtained by comparing the reaction rates as determined from the counting data from the unknown and reference field, with proper corrections to account for spectral differences between the two fields (see Section 5). One important feature of this technique is that it eliminates the need for knowing the detector efficiency.1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-temperature environments are discussed in 9.2. For those circumstances where the use of cadmium as a thermal shield is undesirable because of potential spectrum perturbations or of temperatures above the melting point of cadmium, the method described in Test Method E 481 can be used in some cases. Alternatively, gadolinium filters may be used instead of cadmium. For high temperature applications in which aluminum alloys are unsuitable, other alloys such as cobalt-nickel or cobalt-vanadium have been used. 1.3 This test method may be used to determine the equivalent 2200 m/s fluence rate. The accurate determination of the actual thermal neutron fluence rate requires knowledge of the neutron temperature, and determination of the neutron temperature is not within the scope of the standard. 1.4 The techniques presented are suitable on......

Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

ASTM C1689-08 六氟化铀子二次采样的标准操作规程

Uranium hexafluoride is normally produced and handled in large (typically 1- to 14-ton) quantities and must, therefore be characterized by reference to representative samples. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designated equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice indicates appropriate principles, equipment and procedures currently in use for subsampling of liquid UF6. It is used by UF6 converters, enrichers and fuel fabricators to review the effectiveness of existing procedures or to design equipment and procedures for future use. Other subsampling procedures such as UF6 vapor sampling are not directly representative of the chemical quality of liquid UF6. It is emphasized that this test guide is not meant to address conventional or nuclear criticality safety issues.1.1 This practice is applicable to subsampling uranium hexafluoride (UF6), using heat liquefaction techniques, from bulk containers, obtained in conformance with C 1052, into smaller sample containers, which are required for laboratory analyses. 1.2 It is assumed that the liquid UF6 being sampled comprises a single quality and quantity of material. This practice does not address any special additional arrangement that might be required for taking proportional or composite samples. 1.3 The number of samples to be taken, their nominal sample weight, and their disposition shall be agreed upon between the parties. 1.4 The scope of this practice does not include provisions for preventing criticality incidents. 1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Practice for Subsampling of Uranium Hexafluoride

ASTM C1553-08 核废料干燥行为的标准指南

This guide provides technical information for use by SNF owners to determine the forms of water usually associated with spent nuclear fuel due to corrosion damage of the fuel, cladding and storage materials during irradiation and in storage pools. Drying may be needed to prepare the SNF for sealed dry storage, transportation, and/or permanent disposal at a repository. This guide provides information for: 4.1.1 Evaluating what drying system should be used, 4.1.2 Drying methods, and 4.1.3 Methods to confirm that adequate dryness was achieved. The guide can be used to determine: 4.2.1 Drying technologies that are designed to remove most of the unbound water but will not remove all forms of water. Water remaining on and in commercial and research reactor spent nuclear fuels coming from water basin storage may become an issue when the fuel is sealed in a dry storage system or transport cask. The movement to a dry storage environment typically results in an increase in fuel temperature due to the decay heat. This temperature change could be significant to cause the release of water remaining in a sealed dry package that may result in container pressurization, fuel retrievability issues, and container corrosion. 4.2.2 A methodology for evaluating drying processes that may not readily remove all forms of water that may be retained in pores in fuel cladding, capillaries, sludge, crud, and thin wetted surface films. Drying techniques are even less successful in removing bound water. Removal of bound water will only occur when the specific threshold energy is applied to break the bonds involved and release the water. For spent nuclear fuel this threshold energy may come from the combination of thermal input and ionizing radiation. 4.2.3 How the residual water retained with the SNF, CRUD and sludge inside a sealed package may become available to react with the internal environment, the fuel, and the package materials as a result of extended time at equilibrium dry storage temperatures, or as the direct result of radiolytic decomposition.1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate dryness may be readily achieved for undamaged commercial fuel but may become a complex issue for any SNF where cladding damage has occurred during fuel irradiation, storage, or both, at the spent fuel pools. Dryness issues may also result from the presence of sludge, crud, and other hydrated compounds connected to the SNF that hold water and resist drying efforts. 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 Guide for Drying Behavior of Spent Nuclear Fuel

ASTM C1689-08a 六氟化铀子二次采样的标准方法

Uranium hexafluoride is normally produced and handled in large (typically 1- to 14-ton) quantities and must, therefore be characterized by reference to representative samples. The samples are used to determine compliance with the applicable commercial Specifications C 996 and C 787 by means of the appropriate test method (for example, Test Method C 761 and references therein). The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designated equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice indicates appropriate principles, equipment and procedures currently in use for subsampling of liquid UF6. It is used by UF6 converters, enrichers and fuel fabricators to review the effectiveness of existing procedures or to design equipment and procedures for future use. Other subsampling procedures such as UF6 vapor sampling are not directly representative of the chemical quality of liquid UF6. It is emphasized that this test guide is not meant to address conventional or nuclear criticality safety issues.1.1 This practice is applicable to subsampling uranium hexafluoride (UF6), using heat liquefaction techniques, from bulk containers, obtained in conformance with C 1052, into smaller sample containers, which are required for laboratory analyses. 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 It is assumed that the liquid UF6 being sampled comprises a single quality and quantity of material. This practice does not address any special additional arrangement that might be required for taking proportional or composite samples. 1.4 The number of samples to be taken, their nominal sample weight, and their disposition shall be agreed upon between the parties. 1.5 The scope of this practice does not include provisions for preventing criticality incidents. 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 Practice for Subsampling of Uranium Hexafluoride

ASTM C1347-08 分析用铀材料制备和溶解的标准实施规程

The materials covered that must meet ASTM specifications are uranium metal and uranium oxide. 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 C 753 and C 776. The material is assayed for uranium to determine whether the content is as specified. 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 TitleSection Dissolution of Uranium Metal and Oxide with Nitric Acid8.1 Dissolution of Uranium Oxides with Nitric Acid and Residue Treatment8.2 Dissolution of Uranium-Aluminum Alloys in Hydrochloric Acid with Residue Treatment8.3 Dissolution of Uranium Scrap and Ash by Leaching with Nitric Acid and Treatment of Residue by Carbonate Fusion8.4 Dissolution of Refractory Uranium-Containing Material by Carbonate Fusion8.5 Dissolution of Uranium—Aluminum Alloys Uranium Scrap and Ash, and Refractory Uranium-Containing Materials by Microwave Treatment8.6 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 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. Specific hazards statements are given in Section 7.

Standard Practice for Preparation and Dissolution of Uranium Materials for Analysis

ASTM C1703-08 气态六氟化铀采样的标准方法

Uranium hexafluoride is normally produced and handled in large (typically 1 to 14-ton) quantities and must, therefore, be characterized by reference to representative samples (see ISO/DIS 7195). The samples are used to determine compliance with the applicable commercial specifications C 996 and C 787. The quantities involved, physical properties, chemical reactivity, and hazardous nature of UF6 are such that for representative sampling, specially designed equipment must be used and operated in accordance with the most carefully controlled and stringent procedures. This practice can be used by UF6 converters, enrichers, and fuel fabricators to review the effectiveness of existing procedures or as a guide to the design of equipment and procedures for future use. The intention of this practice is to avoid liquid UF6 sampling once the cylinder has been filled. For safety reasons, manipulation of large quantities of liquid UF6 should be avoided when possible. It is emphasized that this practice is not meant to address conventional or nuclear criticality safety issues.1.1 This practice covers methods for withdrawing representative sample(s) of uranium hexafluoride (UF6) during a transfer occurring in the gas phase. Such transfer in the gas phase can take place from a mother cylinder, for example in an autoclave to a receiving cylinder. It can also occur during the filling in the gas phase of a cylinder during a continuous production process, for example centrifuge enrichment facility or the distillation column in a conversion facility. Such sample(s) may be used for determining compliance with the applicable commercial specification, for example Specification C 996 or Specification C 787. 1.2 Since UF6 sampling is taken during the filling process, this practice does not address any special additional arrangements that may be agreed upon between the buyer and the seller when the sampled bulk material is being added to residues already present in a container (“heels recycle”). Such arrangements will be based on QA procedures such as traceability of cylinder origin (to prevent for example contamination with irradiated material). 1.3 If the receiving cylinder is purged after filling and sampling, special verifications must be performed by the user to verify the representativity of the sample(s). It is then expected that the results found on volatile impurities with gas phase sampling may be conservative. 1.4 This practice is only applicable when the transfer occurs in the gas phase. When the transfer is performed in the liquid phase, Practice C 1052 should apply. This practice does not apply to gas sampling after the cylinder has been filled since the sample taken will not be representative of the cylinder. 1.5 The scope of this practice does not include provisions for preventing criticality incidents. 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 Practice for Sampling of Gaseous Uranium Hexafluoride