This document specifies a method for the determination of the isotopic and elemental uranium and plutonium concentrations of nuclear materials in nitric acid solutions by thermal-ionization mass spectrometry.
The method applies to uranium and plutonium isotope composition and concentration measurement of irradiated Magnox and light water reactor fuels (boiling water reactor or pressurized water reactor), in final products at spent-fuel reprocessing plants, and in feed and products of MOX and uranium fuel fabrication. The method is applicable to other fuels, but the chemical separation and spike solution are, if necessary, adapted to suit each type of fuel.

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This document specifies the dissolution of samples consisting of MOX pellets or powders to provide suitable aliquots for subsequent analysis of elemental concentration and isotopic composition.

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This document specifies an analytical method by spectrophotometry, for determining the plutonium concentration in nitric acid solutions, with spectrophotometer implemented in hot cell and glove box allowing the analysis of high activity solutions. Commonly, the method is applicable, without interference, even in the presence of numerous cations, for a plutonium concentration higher than 0,5 mg·l−1 in the original sample with a standard uncertainty, with coverage factor k = 1, less than 5 %.
The method is intended for process controls at the different steps of the process in a nuclear fuel reprocessing plant or in other nuclear facilities.

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This document specifies the dissolution of powder samples of plutonium oxide for subsequent determination of elemental concentration and isotopic composition.

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This document describes the ceramographic preparation of uranium dioxide (UO2) sintered pellets for qualitative and quantitative microstructure examinations.
These examinations can be carried out before and after thermal or chemical etching.
They enable
— observations of fissures, inter- or intra-granular pores and inclusions, and
— measurement of pore and grain size and measurement of pore and grain size distributions.
The measurement of average grain size can be carried out using a classical counting method as described in ISO 2624 or ASTM E112[3], i.e. intercept procedure, comparison with standard grids or reference photographs.
The measurement of pore-size distributions is usually carried out by an automatic image analyser. If the grain-size distributions are also measured with an image analyser, it is recommended that thermal etching be used to reveal the grain structure uniformly throughout the whole sample.

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This document specifies a method of determining the apparent density and tap density of free-flowing uranium dioxide (UO2) powder which will be used for pelleting and sintering of UO2 pellets as a nuclear fuel.
This method can be used for different UO2 powder types including grains, granules, spheres or other kinds of particles. The method can also be applied to other fuel powders as PuO2, ThO2 and powder mixtures as UO2-PuO2 and UO2-Gd2O3.
This document is based on the principle of using a flowmeter funnel (see 4.1). Other measurement apparatus, such as a Scott volumeter, can also be used.

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This document specifies gas leakage test criteria and test methods for demonstrating that packages used to transport radioactive materials comply with the package containment requirements defined in the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material for:
— design verification;
— fabrication verification;
— preshipment verification;
— periodic verification;
— maintenance verification.
This document describes a method for relating permissible activity release of the radioactive contents carried within a containment system to equivalent gas leakage rates under specified test conditions. This approach is called gas leakage test methodology. However, in this document it is recognized that other methodologies might be acceptable, provided that they demonstrate that any release of the radioactive contents will not exceed the regulatory requirements, and subject to agreement with the competent authority.
This document provides both overall and detailed guidance on the complex relationships between an equivalent gas leakage test and a permissible activity release rate. Whereas the overall guidance is universally agreed upon, the use of the detailed guidance shall be agreed upon with the competent authority during the Type B(U), Type B(M) or Type C packages certification process.
It should be noted that, for a given package, demonstration of compliance is not limited to a single methodology.
While this document does not require particular gas leakage test procedures, it does present minimum requirements for any test that is to be used. It is the responsibility of the package designer or consignor to estimate or determine the maximum permissible release rate of radioactivity to the environment and to select appropriate leakage test procedures that have adequate sensitivity.
This document pertains specifically to Type B(U), Type B(M) or Type C packages for which the regulatory containment requirements are specified explicitly.

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This document provides the following: — specifications for cylinders for the transport of uranium hexafluoride (UF6) to provide compatibility among different users, — description of cylinder designs, but is not intended to develop new designs, — fabrication requirements for the procurement of new cylinders designed for the transport of 0,1 kg or more of uranium hexafluoride, — fabrication requirements for the procurement of new valve protections, valves and plugs, and — requirements for cylinders and valve protections in service.

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ISO 18557 presents guidelines for sampling strategies and characterization processes to assess the contamination of soils, buildings and infrastructures, prior to remediation and/or to check that the remediation objectives have been met (final release surveys). The principles presented need to be appropriately graded as regards the specific situations concerned (size, level of contamination?). It can be used in conjunction with each country's key documentation.
ISO 18557 deals with characterization in relation to site remediation. It applies to sites contaminated after normal operation of older nuclear facilities. It could also apply to site remediation after a major accident, and in this case the input data will be linked to the accident involved.
ISO 18557 complements existing standards, notably concerning sampling, sample preservation and their transport, treatment and laboratory measurements, but also those related to in situ chemical and radiological measurements. References in the Bibliography contain links to appropriate documentation and techniques as required by individual member countries.
ISO 18557 does not apply to the following issues: execution of clean-up works, sampling and characterization of waste (conditioned or unconditioned) or to waste packages.
It does not apply to groundwater characterization (saturated zone).
Given the case-by-case nature of site remediation and decommissioning, the principles and guidance communicated in ISO 18557 are intended as general guidance only, not prescriptive requirements.

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This document provides specific requirements and guidance on the nuclear criticality safety of waste containing fissile nuclides, generated during normal operations. This document is intended to be used along-side and in addition to ISO 1709. This document applies specifically to the nuclear criticality safety of solid nuclear wastes. It also applies to residual quantities of liquids and/or slurries which are either intimately associated with the solid nuclear waste materials or arise as a result of processing or handling the waste. This document does not apply to bulk or process liquids (including higher concentration process solutions) or irradiated or un-irradiated fuel elements. NOTE The term fuel element is sometimes applied to fuel assembly, fuel bundle, fuel cluster, fuel rod, fuel plate, etc. All these terms are based on one or more fuel elements. A cylindrical fuel rod (sometimes referred to as a fuel pin) for a light-water-reactor is an example of a fuel element. All stages of the waste life cycle are within the scope of the document. This document can also be applied to the transport of solid nuclear waste outside the boundaries of nuclear establishments.

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This document covers trunnion systems used for tie-down, tilting and/or lifting of a package of radioactive material during transport operations. Aspects included are the design, manufacture, maintenance, inspection and management system. Regulations which can apply during handling operation in nuclear facilities are not addressed in document. This document does not supersede any of the requirements of international or national regulations, concerning trunnions used for lifting and tie-down.

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This document provides guidance, requirements and recommendations related to determination of limits on subcriticality dimensions and to their compliance with: — geometrical dimensions specified in the design (design dimensions), or, — actual dimensions. This document is applicable to nuclear facilities containing fissile materials, except nuclear power reactor cores. This document can also be applied to the transport of fissile materials outside the boundaries of nuclear establishments. Subcriticality dimension control based on dimensions and layout of fuel assembly, fuel rods and fuel pellets are not covered by this document. This document does not specify requirements related to the control of fissile and non-fissile material compositions. The Quality Assurance associated with the fabrication and layout of the unit based on specifications (e.g. drawings elaborated during design) is a prerequisite of this document. The Quality Assurance is important to ensure the consistency between the unit geometry, its general purpose and its intended function.

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Method for determining the Oxygen-to-Metal (O/M) ratio in mixed uranium-plutonium oxide (U,Pu)O2 ± X pellets. The parameters given in the following paragraphs are relevant for pellets within a range of O/M ratio corresponding to 1,98 to 2,01. The method described in the document is adapted, with regard to the parameters, if the expected values of O/M ratio are outside the range.

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ISO 12183:2016 describes an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, K = 2). The method is suitable for aqueous solutions containing more than 0,5 g/L plutonium and test samples containing between 4 mg and 15 mg of plutonium. Application of this technique to solutions containing less than 0,5 g/L and test samples containing less than 4 mg of plutonium requires experimental demonstration by the user that applicable data quality objectives will be met.
For some applications, purification of test samples by anion exchange is required before measurement to remove interfering substances when present in significant amounts.

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ISO 12800:2017 gives guidelines on the determination of the specific surface area of as-fabricated uranium dioxide powder by volumetric or gravimetric determination of the amount of nitrogen adsorbed on the powder, and can be applied to other similar materials, e.g. U3O8, UO2-PuO2 powders, and other bodies with similar surface areas, e.g. powder granules or green pellets, provided that the conditions described are fulfilled. Modifications using other adsorbing gases are included.
The method is relevant as long as the expected value is in the range between 1 m2/g and 10 m2/g.

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The ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the pellet microstructure.
The examinations are performed before and after thermal treatment or chemical etching.
They allow
- observation of any cracks, intra- and intergranular pores or inclusions, and
- measurement of the grain size, porosity and plutonium homogeneity distribution.
The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2] The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen.
The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching but alpha-autoradiography can also be used. The first technique avoids the tendency for autoradiography to exaggerate the size of plutonium-rich clusters due to the distance the alpha particles travel away from the source.

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ISO 12799:2015 describes a procedure for measuring the nitrogen content of UO2, (U,Gd)O2, and (U,Pu)O2 pellets. Nitrogen in nuclear fuel may be present either as elemental nitrogen or chemically combined in the form of nitrogenous compounds. The technique described herein serves to determine the total content of nitrogen excluding those compounds whose decomposition temperature is above 2 200 °C (most notably Pu and U nitrides).

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Method for determining the Oxygen-to-Metal (O/M) ratio in mixed uranium-plutonium oxide (U,Pu)O2 ± X pellets. The parameters given in the following paragraphs are relevant for pellets within a range of O/M ratio corresponding to 1,98 to 2,01. The method described in the document is adapted, with regard to the parameters, if the expected values of O/M ratio are outside the range.

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ISO 12799:2015 describes a procedure for measuring the nitrogen content of UO2, (U,Gd)O2, and (U,Pu)O2 pellets. Nitrogen in nuclear fuel may be present either as elemental nitrogen or chemically combined in the form of nitrogenous compounds. The technique described herein serves to determine the total content of nitrogen excluding those compounds whose decomposition temperature is above 2 200 °C (most notably Pu and U nitrides).

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The ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the pellet microstructure.
The examinations are performed before and after thermal treatment or chemical etching.
They allow
- observation of any cracks, intra- and intergranular pores or inclusions, and
- measurement of the grain size, porosity and plutonium homogeneity distribution.
The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2] The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen.
The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching but alpha-autoradiography can also be used. The first technique avoids the tendency for autoradiography to exaggerate the size of plutonium-rich clusters due to the distance the alpha particles travel away from the source.

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ISO 12183:2016 describes an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, K = 2). The method is suitable for aqueous solutions containing more than 0,5 g/L plutonium and test samples containing between 4 mg and 15 mg of plutonium. Application of this technique to solutions containing less than 0,5 g/L and test samples containing less than 4 mg of plutonium requires experimental demonstration by the user that applicable data quality objectives will be met.
For some applications, purification of test samples by anion exchange is required before measurement to remove interfering substances when present in significant amounts.

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ISO 12800:2017 gives guidelines on the determination of the specific surface area of as-fabricated uranium dioxide powder by volumetric or gravimetric determination of the amount of nitrogen adsorbed on the powder, and can be applied to other similar materials, e.g. U3O8, UO2-PuO2 powders, and other bodies with similar surface areas, e.g. powder granules or green pellets, provided that the conditions described are fulfilled. Modifications using other adsorbing gases are included.
The method is relevant as long as the expected value is in the range between 1 m2/g and 10 m2/g.

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This document specifies a method of determining the apparent density and tap density of free-flowing uranium dioxide (UO2) powder which will be used for pelleting and sintering of UO2 pellets as a nuclear fuel. This method can be used for different UO2 powder types including grains, granules, spheres or other kinds of particles. The method can also be applied to other fuel powders as PuO2, ThO2 and powder mixtures as UO2-PuO2 and UO2-Gd2O3. This document is based on the principle of using a flowmeter funnel (see 4.1). Other measurement apparatus, such as a Scott volumeter, can also be used.

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This document specifies an analytical method by spectrophotometry, for determining the plutonium concentration in nitric acid solutions, with spectrophotometer implemented in hot cell and glove box allowing the analysis of high activity solutions. Commonly, the method is applicable, without interference, even in the presence of numerous cations, for a plutonium concentration higher than 0,5 mg·l−1 in the original sample with a standard uncertainty, with coverage factor k = 1, less than 5 %. The method is intended for process controls at the different steps of the process in a nuclear fuel reprocessing plant or in other nuclear facilities.

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This document specifies a method for the determination of the isotopic and elemental uranium and plutonium concentrations of nuclear materials in nitric acid solutions by thermal-ionization mass spectrometry. The method applies to uranium and plutonium isotope composition and concentration measurement of irradiated Magnox and light water reactor fuels (boiling water reactor or pressurized water reactor), in final products at spent-fuel reprocessing plants, and in feed and products of MOX and uranium fuel fabrication. The method is applicable to other fuels, but the chemical separation and spike solution are, if necessary, adapted to suit each type of fuel.

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This document specifies the dissolution of powder samples of plutonium oxide for subsequent determination of elemental concentration and isotopic composition.

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This document specifies the dissolution of samples consisting of MOX pellets or powders to provide suitable aliquots for subsequent analysis of elemental concentration and isotopic composition.

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This document provides a method for evaluation of the measurement uncertainty arising when an impurity content of uranium solution is determined by a regression line that has been fitted by the "method of least squares". It is intended to be used by chemical analyzers. Simple linear regression, hereinafter called "basic regression", is defined as a model with a single independent variable that is applied to fit a regression line through n different data points (xi, yi) (i = 1,?, n) in such a way that makes the sum of squared errors, i.e. the squared vertical distances between the data points and the fitted line, as small as possible. For the linear calibration, "classical regression" or "inverse regression" is usually used; however, they are not convenient. Instead, "reversed inverse regression" is used in this document[2]. Reversed inverse regression treats y (the reference solutions) as the response and x (the observed measurements) as the inputs; these values are used to fit a regression line of y on x by the method of least squares. This regression is distinguished from basic regression in that the xi's (i = 1,?, n) vary according to normal distributions but the yi's (i = 1,?, n) are fixed; in basic regression, the yi's vary but the xi's are fixed. The regression line fitting, calculation of combined uncertainty, calculation of effective degrees of freedom, calculation of expanded uncertainty, reflection of reference solutions' uncertainties in the evaluation result, and bias correction are explained in order of mention. Annex A presents a practical example of uncertainty evaluation. Annex B provides a flowchart showing the steps for uncertainty evaluation. In addition, Annex C explains the use of weighting factors for handling non-uniform variances in reversed inverse regression. NOTE 1 In the case of classical regression, the fitted regression line is inverted prior to actual sample measurement[3]. In the case of inverse regression, the roles of x and y are not consistent with the convention that the variable x represents the inputs, whereas the variable y represents the response. For these reasons, the two regressions are excluded from this document. NOTE 2 The term "reversed inverse regression" was suggested taking into account the history of regression analysis theory. Instead, it can be desirable to use some other term, e.g. "pseudo-basic regression".

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This document specifies gas leakage test criteria and test methods for demonstrating that packages used to transport radioactive materials comply with the package containment requirements defined in the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material for: — design verification; — fabrication verification; — preshipment verification; — periodic verification; — maintenance verification. This document describes a method for relating permissible activity release of the radioactive contents carried within a containment system to equivalent gas leakage rates under specified test conditions. This approach is called gas leakage test methodology. However, in this document it is recognized that other methodologies might be acceptable, provided that they demonstrate that any release of the radioactive contents will not exceed the regulatory requirements, and subject to agreement with the competent authority. This document provides both overall and detailed guidance on the complex relationships between an equivalent gas leakage test and a permissible activity release rate. Whereas the overall guidance is universally agreed upon, the use of the detailed guidance shall be agreed upon with the competent authority during the Type B(U), Type B(M) or Type C packages certification process. It should be noted that, for a given package, demonstration of compliance is not limited to a single methodology. While this document does not require particular gas leakage test procedures, it does present minimum requirements for any test that is to be used. It is the responsibility of the package designer or consignor to estimate or determine the maximum permissible release rate of radioactivity to the environment and to select appropriate leakage test procedures that have adequate sensitivity. This document pertains specifically to Type B(U), Type B(M) or Type C packages for which the regulatory containment requirements are specified explicitly.

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This document describes the ceramographic preparation of uranium dioxide (UO2) sintered pellets for qualitative and quantitative microstructure examinations. These examinations can be carried out before and after thermal or chemical etching. They enable — observations of fissures, inter- or intra-granular pores and inclusions, and — measurement of pore and grain size and measurement of pore and grain size distributions. The measurement of average grain size can be carried out using a classical counting method as described in ISO 2624 or ASTM E112[3], i.e. intercept procedure, comparison with standard grids or reference photographs. The measurement of pore-size distributions is usually carried out by an automatic image analyser. If the grain-size distributions are also measured with an image analyser, it is recommended that thermal etching be used to reveal the grain structure uniformly throughout the whole sample.

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ISO 1709:2018 specifies the basic principles and limitations which govern operations with fissile materials. It discusses general nuclear criticality safety criteria for equipment design and for the development of operating controls, while providing guidance for the assessment of procedures, equipment, and operations. It does not cover specific quality assurance requirements or details of equipment or operational procedures. ISO 1709:2018 does not deal with the issues associated with administrative criteria relating to nuclear criticality safety. These issues are covered by ISO 14943. It does not cover the effects of radiation on man or materials, unless the material properties affect nuclear criticality safety. These criteria apply to operations with fissile materials outside nuclear reactors but within the boundaries of nuclear establishments. They are concerned with the limitations which are imposed on operations because of the properties of these materials which permit them to support nuclear chain reactions. These principles apply to quantities of fissile nuclides in which nuclear criticality safety is required to be established. ISO 1709:2018 can also be applied to the transport of fissile materials outside the boundaries of nuclear establishments.

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ISO 21613:2015 describes a method for determining chlorine and fluorine in mixed (U,Pu)O2 powders and sintered pellets. It is applicable for the analysis of samples containing 5 µg.g−1 to 50 µg.g−1 of chlorine and 2 µg.g−1 to 50 µg.g−1of fluorine.
For UO2 powder and sintered pellets, refer to ISO 22875.

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ISO 16424:2012 is applicable to the evaluation of the homogeneity of Gd distribution within gadolinium fuel blends, and the determination of the Gd2O3 content in sintered fuel pellets of Gd2O3+UO2 from 1 % to 10 %, by measurements of gadolinium (Gd) and uranium (U) elements using ICP-AES.
After performing measurements of Gd and U elements using ICP-AES, if statistical methodology is additionally applied, homogeneity of Gd distribution within a Gd fuel pellet lot can also be evaluated. However, ISO 16424:2012 covers the statistical methodology only on a limited basis.

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ISO 15651:2015 describes a procedure for measuring the total hydrogen content of UO2 and PuO2 powders (up to 2 000 µg/g oxide) and of U02 and (U,Gd)O2 and (U,Pu)O2 pellets (up to 10 µg/g oxide). The total hydrogen content results from adsorbed water, water of crystallization, hydro-carbon, and other hydrogenated compounds which can exist as impurities in the fuel.

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ISO 19017:2015 is applicable to gamma radiation measurements on radioactive waste.
Radioactive waste can be found in different forms and exhibit a wide range of characteristics, including the following:
- raw or unconditioned waste, including process waste (filters, resins, control rods, scrap, etc.) and waste from dismantling or decommissioning;
- conditioned waste in various forms and matrices (bitumen, cement, hydraulic binder, etc.);
- very low level (VLLW), low level (LLW), intermediate level (ILW) and high level radioactive waste (HLW);
- different package shapes: cylinders, cubes, parallelepipeds, etc.
Guidance is provided in respect of implementation, calibration, and quality control. The diversity of applications and system realizations (ranging from research to industrial systems, from very low level to high level radioactive waste, from small to large volume packages with different shapes, with different performance requirements and allowable measuring time) renders it impossible to provide specific guidance for all instances; the objective of this International Standard is, therefore, to establish a set of guiding principles. Ultimately, implementation is to be performed by suitably qualified and experienced persons and based on a thorough understanding of the influencing factors, contributing variables and performance requirements of the specific measurement application.
This International Standard assumes that the need for the provision of such a system will have been adequately considered and that its application and performance requirements will have been adequately defined through the use of a structured requirements capture process, such as data quality objectives (DQO).
It is noted that, while outside the scope of this International Standard, many of the principles, measurement methods, and recommended practices discussed here are also equally applicable to gamma measurements of items other than radioactive waste (e.g. bulk food, water, free-standing piles of materials) or to measurements made on radioactive materials contained within non-traditional packages (e.g. in transport containers).

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ISO 21483:2013 specifies an analytical method for determining the solubility in nitric acid of plutonium in pellets of unirradiated mixed oxide fuel (light-water reactor fuels). The results provide information about the expected dissolution behaviour of irradiated pellets under industrial reprocessing conditions. In this aspect, the specific conditions (e.g. time of the test) may vary depending upon the need to match to a specific reprocessor's requirements. The test is aimed at determining solubility under equilibrium conditions rather than the kinetics of dissolution and hence allows for a second dissolution period.

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ISO 16424:2012 is applicable to the evaluation of the homogeneity of Gd distribution within gadolinium fuel blends, and the determination of the Gd2O3 content in sintered fuel pellets of Gd2O3+UO2 from 1 % to 10 %, by measurements of gadolinium (Gd) and uranium (U) elements using ICP-AES.
After performing measurements of Gd and U elements using ICP-AES, if statistical methodology is additionally applied, homogeneity of Gd distribution within a Gd fuel pellet lot can also be evaluated. However, ISO 16424:2012 covers the statistical methodology only on a limited basis.

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ISO 21613:2015 describes a method for determining chlorine and fluorine in mixed (U,Pu)O2 powders and sintered pellets. It is applicable for the analysis of samples containing 5 µg.g−1 to 50 µg.g−1 of chlorine and 2 µg.g−1 to 50 µg.g−1of fluorine.
For UO2 powder and sintered pellets, refer to ISO 22875.

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ISO 15651:2015 describes a procedure for measuring the total hydrogen content of UO2 and PuO2 powders (up to 2 000 µg/g oxide) and of U02 and (U,Gd)O2 and (U,Pu)O2 pellets (up to 10 µg/g oxide). The total hydrogen content results from adsorbed water, water of crystallization, hydro-carbon, and other hydrogenated compounds which can exist as impurities in the fuel.

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ISO 19017:2015 is applicable to gamma radiation measurements on radioactive waste.
Radioactive waste can be found in different forms and exhibit a wide range of characteristics, including the following:
- raw or unconditioned waste, including process waste (filters, resins, control rods, scrap, etc.) and waste from dismantling or decommissioning;
- conditioned waste in various forms and matrices (bitumen, cement, hydraulic binder, etc.);
- very low level (VLLW), low level (LLW), intermediate level (ILW) and high level radioactive waste (HLW);
- different package shapes: cylinders, cubes, parallelepipeds, etc.
Guidance is provided in respect of implementation, calibration, and quality control. The diversity of applications and system realizations (ranging from research to industrial systems, from very low level to high level radioactive waste, from small to large volume packages with different shapes, with different performance requirements and allowable measuring time) renders it impossible to provide specific guidance for all instances; the objective of this International Standard is, therefore, to establish a set of guiding principles. Ultimately, implementation is to be performed by suitably qualified and experienced persons and based on a thorough understanding of the influencing factors, contributing variables and performance requirements of the specific measurement application.
This International Standard assumes that the need for the provision of such a system will have been adequately considered and that its application and performance requirements will have been adequately defined through the use of a structured requirements capture process, such as data quality objectives (DQO).
It is noted that, while outside the scope of this International Standard, many of the principles, measurement methods, and recommended practices discussed here are also equally applicable to gamma measurements of items other than radioactive waste (e.g. bulk food, water, free-standing piles of materials) or to measurements made on radioactive materials contained within non-traditional packages (e.g. in transport containers).

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ISO 21483:2013 specifies an analytical method for determining the solubility in nitric acid of plutonium in pellets of unirradiated mixed oxide fuel (light-water reactor fuels). The results provide information about the expected dissolution behaviour of irradiated pellets under industrial reprocessing conditions. In this aspect, the specific conditions (e.g. time of the test) may vary depending upon the need to match to a specific reprocessor's requirements. The test is aimed at determining solubility under equilibrium conditions rather than the kinetics of dissolution and hence allows for a second dissolution period.

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ISO 18557 presents guidelines for sampling strategies and characterization processes to assess the contamination of soils, buildings and infrastructures, prior to remediation and/or to check that the remediation objectives have been met (final release surveys). The principles presented need to be appropriately graded as regards the specific situations concerned (size, level of contamination?). It can be used in conjunction with each country's key documentation. ISO 18557 deals with characterization in relation to site remediation. It applies to sites contaminated after normal operation of older nuclear facilities. It could also apply to site remediation after a major accident, and in this case the input data will be linked to the accident involved. ISO 18557 complements existing standards, notably concerning sampling, sample preservation and their transport, treatment and laboratory measurements, but also those related to in situ chemical and radiological measurements. References in the Bibliography contain links to appropriate documentation and techniques as required by individual member countries. ISO 18557 does not apply to the following issues: execution of clean-up works, sampling and characterization of waste (conditioned or unconditioned) or to waste packages. It does not apply to groundwater characterization (saturated zone). Given the case-by-case nature of site remediation and decommissioning, the principles and guidance communicated in ISO 18557 are intended as general guidance only, not prescriptive requirements.

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ISO 22875:2017 describes a method for determining chlorine and fluorine in uranium dioxide powder and sintered pellets. It is applicable for the measurement of samples with a mass fraction of chlorine from 5 µg/g to 500 µg/g and with a mass fraction of fluorine from 2 µg/g to 500 µg/g.

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ISO 12800:2017 gives guidelines on the determination of the specific surface area of as-fabricated uranium dioxide powder by volumetric or gravimetric determination of the amount of nitrogen adsorbed on the powder, and can be applied to other similar materials, e.g. U3O8, UO2-PuO2 powders, and other bodies with similar surface areas, e.g. powder granules or green pellets, provided that the conditions described are fulfilled. Modifications using other adsorbing gases are included. The method is relevant as long as the expected value is in the range between 1 m2/g and 10 m2/g.

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ISO 21484:2017 describes a method for determining the Oxygen-to-Metal (O/M) ratio in mixed uranium-plutonium oxide (U,Pu)O2 ± X pellets. The parameters given in the following paragraphs are relevant for pellets within a range of O/M ratio corresponding to 1,98 to 2,01. The method described in the document is adapted, with regard to the parameters, if the expected values of O/M ratio are outside the range.

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ISO 22765:2016 describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the pellet microstructure. The examinations are performed before and after thermal treatment or chemical etching. They allow - observation of any cracks, intra- and intergranular pores or inclusions, and - measurement of the grain size, porosity and plutonium homogeneity distribution. The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2] The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen. The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching but alpha-autoradiography can also be used. The first technique avoids the tendency for autoradiography to exaggerate the size of plutonium-rich clusters due to the distance the alpha particles travel away from the source.

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ISO 12183:2016 describes an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, K = 2). The method is suitable for aqueous solutions containing more than 0,5 g/L plutonium and test samples containing between 4 mg and 15 mg of plutonium. Application of this technique to solutions containing less than 0,5 g/L and test samples containing less than 4 mg of plutonium requires experimental demonstration by the user that applicable data quality objectives will be met. For some applications, purification of test samples by anion exchange is required before measurement to remove interfering substances when present in significant amounts.

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ISO 15366-2:2014 describes procedures to chemically separate and purify uranium and plutonium in dissolved solutions of irradiated light water reactor fuels and in samples of high active liquid waste of spent fuel reprocessing plants, prior to their isotopic analysis by e.g. mass spectrometric method or alpha spectrometry. ISO 15366-2:2014 describes a slightly different separation technique from ISO 15366-1, based on the same chemistry, using smaller columns, different support material and special purification steps, applicable to samples containing plutonium and uranium amounts in the nanogram range and below. The detection limits were found to be 500 pg plutonium and 500 pg uranium.

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ISO 15646:2014 describes a procedure for measuring the densification of UO2, (U,Gd)O2, and (U,Pu)O2 pellets, achieved by heat treatment under defined conditions.
The densification of fuel in power operation is an important design feature. Essentially, it is dependent on structural parameters such as pore size, spatial pore distribution, grain size, and in the case of (U,Gd)O2 and (U,Pu)O2, oxide phase structure. A thermal re-sintering test can be used to characterize the dimensional behaviour of the pellets under high temperature. The results of this test are used by the fuel designer to predict dimensional behaviour in the reactor, because thermal densification in the reactor is also dependent on these structural parameters, albeit in a differing manner in terms of quantity.

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ISO 15366-1:2014 describes procedures to chemically separate and purify uranium and plutonium in dissolved solutions of irradiated light water reactor fuels and in samples of high active liquid waste of spent fuel reprocessing plants, prior to their isotopic analysis by e.g. mass spectrometric method or alpha spectrometry. ISO 15366-1:2014 describes a technique for the separation of uranium and plutonium in spent fuel type samples based on chromatographic method. The procedure applies to samples containing 1 μg to 150 μg Pu (IV) and (VI) and 0,1 mg to 2 mg U (IV) and (VI) in up to 2 ml of 3 mol·l-1 nitric acid solution. It is applicable to mixtures of uranium and plutonium in which the U/Pu-ratio can range from 0 up to 200.

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