F46 核材料、核燃料及其分析试验方法 标准查询与下载

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 E526-08 用钛的放射性测定快中子反应率的标准试验方法

Refer to Guide E 844 for the selection, irradiation, and quality control of neutron dosimeters. Refer to Test Method E 261 for a general discussion of the determination of fast-neutron fluence rate with threshold detectors. Titanium has good physical strength, is easily fabricated, has excellent corrosion resistance, has a melting temperature of 1675°C, and can be obtained with satisfactory purity. 46Sc has a half-life of 83.79 days. The 46Sc decay emits a 0.8893 MeV gamma 99.984 % of the time and a second gamma with an energy of 1.1205 MeV 99.987 % of the time. The isotopic content of natural titanium recommended for 46Ti is 8.25 %. The radioactive products of the neutron reactions 47Ti(n,p)47Sc (τ1/2 = 3.3492 d) and 48Ti(n,p)48Sc (τ1/2 = 43.67 h), might interfere with the analysis of 46Sc. Contaminant activities (for example, 65Zn and 182Ta) might interfere with the analysis of 46Sc. See Sections 7.1.2 and 7.1.3 for more details on the 182Ta and 65Zn interference. 46Ti and 46Sc have cross sections for thermal neutrons of 0.59 and 8 barns, respectively ; therefore, when an irradiation exceeds a thermal-neutron fluence greater than about 2 × 1021 cm–2, provisions should be made to either use a thermal-neutron shield to prevent burn-up of 46Sc or measure the thermal-neutron fluence rate and calculate the burn-up. Fig. 1 shows a plot of cross section versus neutron energy for the fast-neutron reactions of titanium which produce 46Sc [that is, NatTi(n,X)46Sc]. Included in the plot is the 46Ti(n,p) reaction and the 47Ti(n,np) contribution to the 46Sc production, normalized (at 14.7 MeV) per 46Ti atom. This figure is for illustrative purposes only to indicate the range of response of the 46Ti(n,p) reaction. Refer to Guide E 1018 for descriptions of recommended tabulated dosimetry cross sections.1.1 This test method covers procedures for measuring reaction rates by the activation reactions 46Ti(n,p) 46Sc + 47Ti(n, np)46Sc. Note 18212;Since the cross section for the (n,np) reaction is relatively small for energies less than 12 MeV and is not easily distinguished from that of the (n,p) reaction, this test method will refer to the (n,p) reaction only. 1.2 The reaction is useful for measuring neutrons with energies above approximately 4.4 MeV and for irradiation times up to about 250 days (for longer irradiations, see Practice E 261). 1.3 With suitable techniques, fission-neutron fluence rates above 109 cm–2·s–1 can be determined. However, in the presence of a high thermal-neutron fluence rate, 46Sc depletion should be investigated. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its......

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

ASTM C967-08 铀矿石浓缩物的标准规范

1.1 This specification covers uranium ore concentrate containing a minimum of 65 mass % uranium. 1.2 This specification does not include requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, national, state, and local regulations pertaining to possessing, shipping, or using source nuclear material (see 2.2). 1.3 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Standard Specification for Uranium Ore Concentrate

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 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 E393-08 通过分析由裂变剂量计产生的钡140来测定反应速率的试验方法

Refer to Guide E 844 for the selection, irradiation, and quality control of neutron dosimeters. Refer to Practice E 261 for a general discussion of the measurement of neutron fluence rate and fluence. The neutron spectrum must be known in order to measure neutron fluence rates with a single detector. Also it is noted that cross sections are continuously being reevaluated. The latest recommended cross sections and details on how they can be obtained are discussed in Guide E 1018. The reaction rate of a detector nuclide of known cross section, when combined with information about the neutron spectrum, permits the determination of the magnitude of the fluence rate impinging on the detector. Furthermore, if results from other detectors are available, the neutron spectrum can be defined more accurately. The techniques for fluence rate and fluence determinations are explained in Practice E 261. 140Ba is a radioactive nuclide formed as a result of uranium fission. Although it is formed in fission of any heavy atom, the relative yield will differ. Recommended fission yields for 140Ba production are given in Table 1. The direct (independent) fission yield of the daughter product 140La, which is counted, is given in Table 2. These independent fission yields are relatively low compared to the 140Ba cumulative fission yield and will not significantly affect the accuracy of the nondestructive procedure and need not be considered. The half-life of 140Ba is 12.752 days. Its daughter 140La has a half-life of 1.6781 days. The comparatively long half-life of 140Ba allows the counting to be delayed several weeks after irradiation in a high-neutron field. However, to achieve maximum sensitivity the daughter product 140La should be counted five to six days after the irradiation during nondestructive analysis or five to six days after chemical separation if the latter technique is used. An alternative method after chemical separation is to count the 140Ba directly. Because of its 12.752 day half-life and substantial fission yield, 140Ba is useful for irradiation times up to about six weeks in moderate intensity fields. The number of fissions produced should be approximately 109 or greater for good counting statistics. Also, if the irradiation time is substantially longer than six weeks, the neutron fluence rate determined will apply mainly to the neutron field existing during the latter part of the irradiation. The 140Ba decay constant and yield are known more accurately than those of many fission products, so it is sometimes used as a standard or base reaction with which other measurements can be normalized.1.1 This test method describes two procedures for the measurement of reaction rates by determining the amount of the fission product 140Ba produced by the non-threshold reactions 235U(n,f), 241Am(n,f), and 239Pu(n,f), and by the threshold reactions 238U(n,f), 237Np(n,f), and 232Th(n,f). 1.2 These reactions produce many fission products, among which is 140Ba, having a half-life of 12.752 days. 140Ba emits gamma rays of several energies; however, these are not easily detected in the presence of other fission products. Competing activity from other fission products requires that a chemical separation be employed or that the 140Ba activity be determined indirectly by counting its daughter product 140La. This test method describes both procedure (a), the nondestructive determination of 140Ba by the direct counting of 140La several days after irradiation, and pro......

Standard Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters

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 E692-08 用高分辨率γ射线光谱分析法对照射过的核燃料中的铯137的标准试验方法

This test method uses a high-resolution gamma-ray spectrometer as a basis for measuring the gamma-ray emission rate of 137Cs-137mBa in a dilute nitric acid solution containing 10 mg/L of cesium carrier. No chemical separation of the cesium from the dissolved-fuel solution is required. The principal steps consist of diluting a weighed aliquot of the dissolved-fuel solution with a known mass of 1 M nitric acid (HNO3) and measuring the 662 keV gamma-ray count rate from the sample, then measuring the 662 keV gamma-ray count rate from a standard source that has the same physical form and counting geometry as the sample. The amount of fuel sample required for the analysis is small. For a sample containing 0.1 g of fuel irradiated to one atom percent fission, a net count rate of approximately 105 counts per second will be observed for a counting geometry that yields a full-energy peak efficiency fraction of 1 × 10-3. The advantage of this small amount of sample is that the concentration of fuel material can be kept at levels well below 1 g/L, which results in negligible self-absorption in the sample aliquot and a small radiation hazard to the analyst.1.1 This test method covers the determination of the number of atoms of 137Cs in aqueous solutions of irradiated uranium and plutonium nuclear fuel. When combined with a method for determining the initial number of fissile atoms in the fuel, the results of this analysis allows atom percent fission (burn-up) to be calculated (1). The determination of atom percent fission, uranium and plutonium concentrations, and isotopic abundances are covered in Test Methods E 267 and E 321. 1.2 137Cs is not suitable as a fission monitor for samples that may have lost cesium during reactor operation. For example, a large temperature gradient enhances 137Cs migration from the fuel region to cooler regions such as the radial fuel-clad gap, or, to a lesser extent, towards the axial fuel end. 1.3 A nonuniform 137Cs distribution should alert the analyst to the potential loss of the fission product nuclide. The 137Cs distribution may be ascertained by an axial gamma-ray scan of the fuel element to be assayed. In a mixed-oxide fuel, comparison of the 137Cs distribution with the distribution of nonmigrating fission-product nuclides such as 95Zr or 144Ce would indicate the relative degree of 137Cs migration. 1.4 The values stated in SI units are to be regarded as standard. No other unites of measurement are included in this 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.

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

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 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 C1514-08 用同位素浓缩度计原理测定铀235微粒的标准试验方法

The enrichment meter principle provides a nondestructive measurement of the 235U fraction of uranium-bearing items. Sampling is not required and no waste is generated, minimizing exposure to hazardous materials and resulting in reduced sampling error. This method relies on a fixed and controlled geometry. The uranium-bearing materials in the measured items and calibration reference materials used for calibration must fill the detector field of view. Use of a low resolution detector (for example, NaI detector) to measure uranium with 235U fraction approximately 10 % which is contained in a thin-walled container can provide a rapid (typically 100 s), easily portable measurement system with precision of 0.6 % and bias of less than 1 %. Use of a high resolution detector (for example, high-purity germanium) can provide measurement with a precision better than 0.2 % and a bias less than 1 % within a 300-s measurement time when measuring uranium with 235U fraction in the range of 0.711 % or above which is contained in thin-walled containers. In order to obtain optimum results using this method, the chemical composition of the item must be well known, the container wall must permit transmission of the 185.7 keV gamma-ray, and the uranium-bearing material within the item must be infinitely thick with respect to the 185.7 keV gamma-ray. All items must be in identical containers or must have a known container wall thickness and composition. Items to be measured must be homogeneous with respect to both 235U fraction and chemical composition. When measuring items, using low-resolution detectors, in thin-walled containers that have not reached secular equilibrium (more than about 120 days after processing), either the method should not be used, additional corrections should be made to account for the age of the uranium, or high-resolution measurements should be performed. The method is often used as a enrichment verification technique.1.1 This test method covers the quantitative determination of the fraction of 235U in uranium using measurement of the 185.7 keV gamma-ray produced during the decay of 235U. 1.2 This test method is applicable to items containing homogeneous uranium-bearing materials of known chemical composition in which the compound is considered infinitely thick with respect to 185.7 keV gamma-rays. 1.3 This test method can be used for the entire range of 235U fraction as a weight percent, from depleted (0.2 % 235U) to highly enriched (97.5 % 235U). 1.4 Measurement of items that have not reached secular equilibrium between 238U and 234Th may not produce the stated bias when low-resolution detectors are used with the computational method listed in Annex A2. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard may involve hazardous materials, operations, and equipment. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Standard Test Method for Measurement of ²³⁵U Fraction Using Enrichment Meter Principle

ASTM C1539-08 液体闪烁计数测定六氟化铀中锝99的标准试验方法

Uranium hexafluoride is a basic material used to prepare nuclear reactor fuel. To be suitable for this purpose, the material must meet the criteria for technetium composition. This test method is designed to determine whether the material meets the requirements described in Specifications C 787 and C 996. Using the specified instrumentation and parameters, this method has a lower detection limit of 0.0004 μgTc/gU. Note 18212;Different instrumentation or parameters may provide varying detection limits, as calculated in 11.4.1.1 This test method is a quantitative method used to determine technetium-99 (99Tc) in uranium hexafluoride (UF6) by liquid scintillation counting. 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 Determination of Technetium-99 in Uranium Hexafluoride by Liquid Scintillation Counting

EJ/T 1221-2007 六氟化铀中微量溴的测定 分光光度法

本标准规定了测定六氟化铀中微量溴的方法原理、试剂、分析步骤、结果计算和方法精密度等。 本标准适用于六氟化铀中微量溴的测定。 取含1g铀的六氟化铀水解液中，方法的测定范围为（2～16）μg/gU。

Determination of micro bromine in uranium hexafluoride by spectrophotometry

EJ/T 1220-2007 天然六氟化铀中金属杂质元素的测定 电感耦合等离子体发射光谱法(ICP-AES法)

本标准规定了 ICP-AES 法测定天然六氟化铀中金属杂质元素的试剂与材料、仪器与设备、试样、分析步骤及精密度和回收率。 本标准适用于天然六氟化铀中银、铝、钡、铍、铋、钙、镉、铜、铁、钾、锂、镁、锰、钠、镍、 铅、锡、锶、锌、锑、钌杂质元素的测定。测定范围见表 1 。 表1 各元素测定范围（略）

Determination of metallic impurities in natural uranium hexafluoride by ICP-AES

EJ/T 1199-2007 重铀酸盐产品常规取样方法

本标准规定了重铀酸盐产品常规取样、制样及水分测定所用的设备和工器具、取样和制样的操作步骤及水分的测定方法。 本标准适用于对重铀酸盐产品进行质量检验时的取样、制样和水分测定。 本标准不适用于产品桶内表层有积水的重铀酸盐产品的取样。

Method of routine sampling for diuranate product

EJ/T 787-2007 烧结二氧化铀芯块外形尺寸、几何密度和外观质量检验方法

本标准规定了烧结二氧化铀芯块外形只寸、几何密度和外观质量检验的仪器设备、试验条件和测量方法等。 本标准适用于无中孔的压水堆烧结二氧化铀芯块（可含钆）外形尺寸、几何密度的测量和外观质量的检验。有中孔的芯块可参照执行。

Test methods for geometric dimension, geometric density and appearance qullity of sintered uranium dioxide pellet

DIN 25712:2007 考虑容器中受辐射轻水反应堆燃料组件运输和储存的燃料燃耗的临界安全

Criticality safety taking into account the burnup of fuel for transport and storage of irradiated light water reactor fuel assemblies in casks

BS ISO 21238:2007 核能.核燃料技术.测定核电站产生的低级和中级放射性废物包放射性的比例因子法

ISO 21238:2007 gives guidelines for the common basic methodology of empirically determining scaling factors to evaluate the radioactivity of difficult-to-measure nuclides in low- and intermediate-level radioactive waste packages. ISO 21238:2007 gives common guidelines for the scaling factors used in the characterization of contaminated wastes produced in nuclear power plants with water-cooled reactor. ISO 21238:2007 is also relevant to other reactor types, such as gas-cooled reactors. Methodologies for determining scaling factors based on theoretical considerations (i.e. not based on experimental measurement) are not covered by ISO 21238:2007.

Nuclear energy. Nuclear fuel technology. Scaling factor method to determine the radioactivity of low and intermediate-level radioactive waste packages generated at nuclear power plants

ISO 9005:2007 核物理.二氧化铀粉末和烧结芯块.用安培法测定氧/铀原子比

This International Standard specifies an analytical method for the determination of the oxygen/uranium atomic ratio in uranium dioxide powder and sintered pellets. The method is applicable to reactor grade samples of hyper-stoichiometric uranium dioxide powder and pellets. The presence of reducing agents or residual organic additives invalidates the procedure.

Nuclear energy - Uranium dioxide powder and sintered pellets - Determination of oxygen/uranium atomic ratio by the amperometric method

ASTM C751-07(2012) 核纯级碳化硼丸的标准规格

1.1 This specification applies to boron carbide pellets for use as a control material in nuclear reactors. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Standard Specification for Nuclear-Grade Boron Carbide Pellets