| Name |
Description |
Abstract |
Status |
Publication date |
Edition |
Number of pages |
Technical committee |
ICS |
| ISO 9279:1992 |
Uranium dioxide pellets — Determination of density and total porosity — Mercury displacement method |
Describes the principle, the apparatus, the procedure, the expression of results and the contents of the test report.
|
Published |
1992-03 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9463:1990 |
Nitric acid feed solutions from reprocessing plants — Spectrophotometric determination of plutonium after oxidation to plutonium(VI) |
The method is applicable, without interference, in the presence of numerous cations, it is applicable to test portions containing between 0,5 mg and 2,5 mg of plutonium. Specifies principle, chemical conditions, reagents, apparatus, operating procedure, expression of results and interferences.
|
Withdrawn |
1990-12 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9463:2009 |
Nuclear energy — Nuclear fuel technology — Determination of plutonium in nitric acid solutions by spectrophotometry |
ISO 9463:2009 specifies an analytical method by spectrophotometry for determining the plutonium concentration of nitric acid solutions in reprocessing plants. The method is applicable, without interference, in the presence of numerous cations, with a standard deviation of about 5 %, where the concentration of plutonium in the solution is at least 0,5 mg·l-1.
|
Withdrawn |
2009-08 |
Edition : 2 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9463:2019 |
Nuclear energy — Nuclear fuel technology — Determination of plutonium in nitric acid solutions by spectrophotometry |
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.
|
Published |
2019-01 |
Edition : 3 |
Number of pages : 12 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9889:1994 |
Determination of carbon content in uranium dioxide powder and sintered pellets — Resistance furnace combustion — Titrimetric/coulometric/infrared absorbtion method |
Specifies a titrimetric/coulometric/infrared absorption method for determining the carbon content in uranium dioxide powder and sintered pellet. Applicable to the determination of 5 µg to 500 µg of carbon in uranium dioxide powder and pellets. Interference from sulfur and halogens is prevented by the use of appropriate traps.
|
Published |
1994-12 |
Edition : 1 |
Number of pages : 9 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9891:1994 |
Determination of carbon content in uranium dioxide powder and sintered pellets — High-frequency induction furnace combustion — Titrimetric/coulometric/infrared absorption methods |
The methods specified are based on heating a portion of the test sample at a temperature of at least 1100 °C to 1200 °C in an oxygen atmosphere, passing the evolved oxidation products over a purification trap filled with manganese dioxide catalyst and silver permanganate catalyst (that oxidises CO to CO2), trapping the CO2, restoring the initial pH continuously by addition of hydroxyl ions either by potentiostatic titrimetry or by coulometry, or alternatively determining the CO2 by IR absorption and integration of the signal obtained. Is applicable to the determination of 5 µg to 500 µg of carbon.
|
Published |
1994-12 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9892:1992 |
Uranium metal, uranium dioxide powder and pellets, and uranyl nitrate solutions — Determination of fluorine content — Fluoride ion selective electrode method |
Specifies an analytical method which can be used within the concentration range of 1 µg to 0,001 g of fluorine per gram of the sample. Specifies the principle, the reagents, the apparatus, the sampling, the procedure, the expression of results and the test report.
|
Published |
1992-04 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 9894:1996 |
Subsampling of uranium hexafluoride in the liquid phase |
Describes a method of subsampling suitable for taking aliquots from a representative sample of uranium hexafluoride in the liquid phase. The subsamples are intended for isotopic analysis (1 g bis 3 g), impurity analysis (10 g bis 200 g) and uranium assay (5 g bis 10 g). Carbon halides, hydrocarbons and certain metal halides can be measured directly from the sample.
|
Published |
1996-06 |
Edition : 1 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 10276:2010 |
Nuclear energy — Fuel technology — Trunnions for packages used to transport radioactive material |
ISO 10276:2010 addresses trunnions fitted to radioactive-material transport packages that are subject to the approval and licensing by competent authorities in accordance with the IAEA No. TS-R-1. Aspects included are design, manufacture, maintenance and quality assurance.
Subject to agreement between the interested parties, ISO 10276:2010 can also be applied to packages that are not subject to the approval by competent authorities.
ISO 10276:2010 covers trunnion systems used for tie-down during transport and trunnions used for tilting and/or lifting.
ISO 10276:2010 does not supersede any of the requirements in the IAEA No. TS-R-1, nor any of the requirements of international or national regulations, concerning trunnions used for lifting and tie-down.
ISO 10276:2010 is applicable to new package design.
|
Withdrawn |
2010-08 |
Edition : 1 |
Number of pages : 20 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 10276:2019 |
Nuclear energy — Fuel technology — Trunnion systems for packages used to transport radioactive material |
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.
|
Published |
2019-12 |
Edition : 2 |
Number of pages : 22 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 10980:1995 |
Validation of the strength of reference solutions used for measuring concentrations |
Describes procedures for preparing reference solutions, procedures for ensuring the quality and the degree of accuracy of the solution strength. Working document for analytical chemists; the underlying statistical theorie is not presented in order to simplify application.
|
Published |
1995-08 |
Edition : 1 |
Number of pages : 42 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 10981:1993 |
Determination of uranium in reprocessing plants dissolver solution — Liquid chromatography method |
The principle of the method specified is diluting the sample in aqueous ascorbic acid solution until the amount of free acid of the nitric solution injected into the chromatograph is lower than 0,1 mol/l in HNO3, performing ion-pair partition chromatography, detecting the elution of uranium by UV spectrophotometry at a wavelength of 254 nm, measuring the peak area, obtaining the result by comparison with the measurements of standards. Applies to uranium concentration between 0,1 g/l and 500 g/l. Specifies how interference by nitrite and plutonium ions is prevented.
|
Withdrawn |
1993-06 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 10981:2004 |
Nuclear fuel technology — Determination of uranium in reprocessing-plant dissolver solution — Liquid chromatography method |
ISO 10981:2004 specifies an analytical method for determining the uranium concentration between 0,1 g/l and 400 g/l in nitric acid solutions of irradiated fuel from light-water reactors, gas-cooled reactors and fast-breeder reactors. It specifies how interference by nitrite and plutonium ions is prevented. The other constituents of fuel solutions do not interfere.
This method is suitable for process control, but not for accountancy purposes.
|
Published |
2004-06 |
Edition : 2 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11311:2011 |
Nuclear criticality safety — Critical values for homogeneous plutonium-uranium oxide fuel mixtures outside of reactors |
ISO 11311:2011 specifies common reference critical values (of which the effective neutron multiplication factor, keff, is equal to 1) for homogeneous water-moderated plutonium-uranium oxide mixtures based on an inter-code comparison of calculated critical values.
It is applicable to operations with unirradiated mixed uranium-plutonium oxide (MOX) outside nuclear reactors.
A classical validation approach for these systems is difficult because of the paucity of critical experiments for MOX fuel.
Various reference systems, in terms of isotopic compositions, thicknesses of water reflection, and densities of oxide are evaluated by different combinations of calculation codes and nuclear data libraries (i.e. different calculation schemes).
The critical values defined in ISO 11311:2011 are the lowest of those calculated by each of these calculation schemes and accepted as credible.
The values in ISO 11311:2011 are reference values and not absolute critical values.
|
Published |
2011-07 |
Edition : 1 |
Number of pages : 13 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11311:2011/Amd 1:2022 |
Nuclear criticality safety — Critical values for homogeneous plutonium-uranium oxide fuel mixtures outside of reactors — Amendment 1: Corrections and clarifications |
|
Published |
2022-12 |
Edition : 1 |
Number of pages : 2 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16795:2004 |
Nuclear energy — Determination of Gd2O3 content of gadolinium fuel pellets by X-ray fluorescence spectrometry |
ISO 16795:2004 covers the determination of Gd2O3 in sintered fuel pellets, by X-ray fluorescence spectrometry using the Gd L-alpha line. The fuel pellets are polished before X-ray examination. This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.
|
Published |
2004-07 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11320:2011 |
Nuclear criticality safety — Emergency preparedness and response |
ISO 11320:2011 provides criteria for emergency preparedness and response to minimize consequences due to a nuclear criticality accident. The criticality safety of operations are evaluated in accordance with ISO 1709.
ISO 11320:2011 applies to a site with one or more facilities which might contain significant quantities and concentrations of fissile material. The extent to which ISO 11320:2011 needs to be applied depends on the overall criticality risk presented by the facilities at the site.
ISO 11320:2011 does not apply to off-site transport and transit storage of packages with fissile material.
ISO 11320:2011 does not apply to sites with operating nuclear power plants or to facilities with research reactors which are licensed to become critical or near-critical, provided that there are no operations with fissile material external to the reactor for which a credible criticality accident risk exists. ISO 11320:2011 can be applied to such sites and facilities in specific cases, if supported by site management and by licensing authorities.
|
Published |
2011-10 |
Edition : 1 |
Number of pages : 9 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11482:1993 |
Guidelines for plutonium dioxide (PuO2) sampling in a nuclear reprocessing plant |
Gives guidelines to ensure that the homogenization and the sampling system supply representative samples from a batch of plutonium dioxide and to ensure that the said sample remains representative until aliquoting. Possible mass change of the sample during transport and storage is also considered. The guidelines apply to PuO2 powder with a specific surface area lower than 25 m^2/g.
|
Published |
1993-12 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11483:1994 |
Preparation of plutonium sources and determination of 238Pu/239Pu isotope ratio by alpha spectrometry |
Describes three simple and fast procedures to prepare plutonium sources and a procedure to measure the activity ratio by alpha particle spectrometry. Does not apply to purified plutonium solutions containing ^241Am nor to those containing more than 10 % of other nonvolatile impurities relative to the Pu content. The methods given are intended for use in conjunction or in parallel with mass spectrometry for the isotopic analysis of plutonium in spent fuel solutions or nuclear grade plutonium products.
|
Withdrawn |
1994-08 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11483:2005 |
Nuclear fuel technology — Preparation of plutonium sources and determination of 238Pu/239Pu isotope ratio by alpha spectrometry |
ISO 11483: 2005 describes three simple and fast procedures to prepare plutonium sources and a procedure to measure the activity ratio of 238Pu to (239Pu + 240Pu) by alpha spectrometry. The alpha spectrometry method is used for the determination of isotopic abundance of 238Pu in combination with mass spectrometry and eliminates the possible interferences of 238U in the latter method. It applies to the analysis of purified solutions of plutonium in 2 mol/l to 4 mol/l nitric acid containing 50 micrograms to 200 micrograms of plutonium per millilitre, as may result from the chemical treatment and purification preceding plutonium isotopic analysis by mass spectrometry.
|
Published |
2005-01 |
Edition : 2 |
Number of pages : 18 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11597:1995 |
Verification of samples of uranyl or plutonium nitrate solutions by density measurements |
Establishes the use of density measurements for verifying the homogeneity of an industrial solution of uranyl or plutonium nitrate.
|
Withdrawn |
1995-07 |
Edition : 1 |
Number of pages : 9 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 11598:1995 |
Requirements for representative sampling of uranyl nitrate solutions for the determination of uranium concentration |
Specifies essential precautions to be taken to ensure that installed tank mixing and sampling systems produce representative samples from a batch of uranyl nitrate solution that remain representative until used.
|
Withdrawn |
1995-07 |
Edition : 1 |
Number of pages : 3 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12183:1995 |
Controlled-potential coulometric assay of plutonium |
Specifies an analytical method for the determination of plutonium in plutonium nitrate solutions of nuclear grade with an accuracy better than 0,1 %. The method is suitable for aqueous solutions containing more than 0,05 g·l^-1 plutonium using test samples between 1 mg and 3 mg of plutonium.
|
Withdrawn |
1995-04 |
Edition : 1 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12183:2005 |
Nuclear fuel technology — Controlled-potential coulometric assay of plutonium |
ISO 12183:2005 specifies an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty of 0,1 % to 0,2 % at the confidence level of 0,95 for a single determination. 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, must be demonstrated by the user as having adequate reliability for their specific application.
|
Withdrawn |
2005-02 |
Edition : 2 |
Number of pages : 24 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO/IEC 80079-20-1:2017/Cor 1:2018 |
Explosive atmospheres — Part 20-1: Material characteristics for gas and vapour classification — Test methods and data — Technical Corrigendum 1 |
|
Published |
2018-07 |
Edition : 1 |
Number of pages : 2 |
Technical Committee |
29.260.20
Electrical apparatus for explosive atmospheres
|
| ISO 12183:2016 |
Nuclear fuel technology — Controlled-potential coulometric assay of plutonium |
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.
|
Published |
2016-08 |
Edition : 3 |
Number of pages : 29 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12184:1994 |
Determination of solubility in nitric acid of plutonium in unirradiated mixed oxide fuel pellets (U,Pu)O2 |
Specifies an analytical method for determining the solubility in nitric acid of plutonium in whole 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.
|
Withdrawn |
1994-08 |
Edition : 1 |
Number of pages : 3 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12795:2004 |
Nuclear fuel technology — Uranium dioxide powder and pellets — Determination of uranium and oxygen/uranium ratio by gravimetric method with impurity correction |
ISO 12795:2004 specifies a method of determining the mass fraction of uranium and the oxygen-to-uranium atomic ratio in hyperstoichiometric uranium dioxide (UO2+X) powders and pellets.
ISO 12795:2004 is used for the determination of the U mass fraction and the O/U-ratio of nuclear grade uranium dioxide.
An impurity correction in the U mass fraction and the O/U-ratio determination should be performed, if the amount of total impurities in oxide form exceeds 1 500 micrograms per gram of sample. Lower impurity levels influence the O/U-ratio by less than 0,000 5 and can be neglected. The non-volatile impurities shall be determined by an appropriate technique, and the correction applied. If the total content of the non-volatile impurities in oxide form is greater than 1 500 micrograms per gram of sample, the overall precision of the method depends on the accuracy of these impurity measurements.
|
Published |
2004-08 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12799:2015 |
Nuclear energy — Determination of nitrogen content in UO2, (U,Gd)O2 and (U,Pu)O2 sintered pellets — Inert gas extraction and conductivity detection method |
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).
|
Published |
2015-03 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12800:2003 |
Nuclear fuel technology — Guide to the measurement of the specific surface area of uranium oxide powders by the BET method |
IS0 12800:2003 covers 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. U308, 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.
|
Withdrawn |
2003-12 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12800:2017 |
Nuclear fuel technology — Guidelines on the measurement of the specific surface area of uranium oxide powders by the BET method |
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.
|
Published |
2017-06 |
Edition : 2 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12803:1997 |
Representative sampling of plutonium nitrate solutions for determination of plutonium concentration |
|
Published |
1997-09 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12807:1996 |
Safe transport of radioactive materials — Leakage testing on packages |
|
Withdrawn |
1996-10 |
Edition : 1 |
Number of pages : 76 |
Technical Committee |
13.280
Radiation protection
;
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21847-3:2007 |
Nuclear fuel technology — Alpha spectrometry — Part 3: Determination of uranium 232 in uranium and its compounds |
ISO 21847-3:2007 describes a method for determining trace amounts of 232U in uranium hexafluoride, uranium oxides or uranyl nitrate.
|
Published |
2007-09 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 12807:2018 |
Safe transport of radioactive materials — Leakage testing on packages |
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.
|
Published |
2018-09 |
Edition : 2 |
Number of pages : 85 |
Technical Committee |
13.280
Radiation protection
;
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 13463:1999 |
Nuclear-grade plutonium dioxide powder for fabrication of light water reactor MOX fuel — Guidelines to help in the definition of a product specification |
|
Published |
1999-09 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 13464:1998 |
Simultaneous determination of uranium and plutonium in dissolver solutions from reprocessing plants — Combined method using K-absorption edge and X-ray fluorescence spectrometry |
|
Published |
1998-04 |
Edition : 1 |
Number of pages : 13 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 13465:1997 |
Determination of neptunium in nitric acid solutions by molecular absorption spectrophotometry |
|
Withdrawn |
1997-05 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 13465:2009 |
Nuclear energy — Nuclear fuel technology — Determination of neptunium in nitric acid solutions by spectrophotometry |
ISO 13465:2009 specifies an analytical method for determining the neptunium concentration by spectrophotometry, with a standard deviation of about 5 %, in nitric acid solutions of nuclear reactor irradiated fuels, at different steps of the process in a nuclear fuel reprocessing plant. The method is applicable to aliquots containing a concentration of neptunium between 10 mg×l-1 and 200 mg×l-1.
|
Published |
2009-05 |
Edition : 2 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15366-1:2014 |
Nuclear fuel technology — Chemical separation and purification of uranium and plutonium in nitric acid solutions for isotopic and isotopic dilution analysis by solvent extraction chromatography — Part 1: Samples containing plutonium in the microgram range and uranium in the milligram range |
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.
|
Published |
2014-07 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15366-2:2014 |
Nuclear fuel technology — Chemical separation and purification of uranium and plutonium in nitric acid solutions for isotopic and isotopic dilution analysis by solvent extraction chromatography — Part 2: Samples containing plutonium and uranium in the nanogram range and below |
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.
|
Published |
2014-07 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15366:1999 |
Nuclear energy — Chemical separation and purification of uranium and plutonium in nitric acid solutions for isotopic and dilution analysis by solvent chromatography |
|
Withdrawn |
1999-10 |
Edition : 1 |
Number of pages : 11 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO/DIS 16795 |
Nuclear Energy — Determination of Gd2O3 content in pellets containing uranium oxide by X-ray fluorescence spectrometry |
ISO 16795:2004 covers the determination of Gd2O3 in sintered fuel pellets, by X-ray fluorescence spectrometry using the Gd L-alpha line. The fuel pellets are polished before X-ray examination. This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.
|
Under development |
|
Edition : 2 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15646:2014 |
Re-sintering test for UO2, (U,Gd)O2 and (U,Pu)O2 pellets |
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.
|
Published |
2014-06 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15647:2004 |
Nuclear energy — Isotopic analysis of uranium hexafluoride — Double-standard gas-source mass spectrometric method |
ISO 15647:2004 specifies a method of isotopic analysis of uranium hexafluoride (UF6) with 235U concentrations less than or equal to 5 % by mass and 234U and 236U concentration between 0,001 % by mass and 1,5 % by mass. The method is in routine use to determine conformance to UF6 specifications.
ISO 15647:2004 is applicable if the mass spectrometer uses Faraday cups.
|
Published |
2004-09 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 15651:2015 |
Nuclear energy — Determination of total hydrogen content in PuO2 and UO2 powders and UO2, (U,Gd)O2 and (U,Pu)O2 sintered pellets — Inert gas extraction and conductivity detection method |
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.
|
Published |
2015-02 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16424:2012 |
Nuclear energy — Evaluation of homogeneity of Gd distribution within gadolinium fuel blends and determination of Gd2O3 content in gadolinium fuel pellets by measurements of uranium and gadolinium elements |
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.
|
Published |
2012-12 |
Edition : 1 |
Number of pages : 15 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16793:2005 |
Nuclear fuel technology — Guide for ceramographic preparation of UO2 sintered pellets for microstructure examination |
ISO 16793:2005 describes the ceramographic preparation of uranium dioxide (UO2) sintered pellets for qualitative and quantitative microstructure examinations.
|
Withdrawn |
2005-02 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16793:2018 |
Nuclear fuel technology — Guidelines for ceramographic preparation of UO2 sintered pellets for microstructure examination |
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.
|
Published |
2018-08 |
Edition : 2 |
Number of pages : 8 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16794:2003 |
Nuclear energy — Determination of carbon compounds and fluorides in uranium hexafluoride infrared spectrometry |
ISO 16794: 2003 specifies a test method for the determination of hydrocarbons, chlorocarbons and partially or completely substituted halocarbons or halohydrocarbons contained as impurities in uranium hexafluoride (UF6) by infrared spectrometry.
This method cannot be used for compounds giving IR rays with interference by UF6 (for example CF4).
The test method is quantitative and applicable in the mole fraction from 0,000 1 % or 0,001 0 %, depending on the type of impurity, up to 0,100 %.
The test method can also be used for the determination of hydrofluoric acid (HF) and several elements existing as fluorides; boron in BF3, silicon in SiF4, phosphorus in PF5, molybdenum in MoF6 and tungsten in WF6.
|
Published |
2003-02 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21847-1:2007 |
Nuclear fuel technology — Alpha spectrometry — Part 1: Determination of neptunium in uranium and its compounds |
ISO 21847-1: 2007 describes a method for determining trace amounts of 237Np in uranium hexafluoride, uranium oxides or uranyl nitrate.
|
Published |
2007-09 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16796:2004 |
Nuclear energy — Determination of Gd2O3 content in gadolinium fuel blends and gadolinium fuel pellets by atomic emission spectrometry using an inductively coupled plasma source (ICP-AES) |
ISO 16796:2004 is applicable to the determination of Gd2O3 in powder blends and sintered pellets of Gd2O3 + UO2 from 1 % to 10 %, by the ICP-AES method.
|
Withdrawn |
2004-08 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16796:2022 |
Nuclear energy — Determination of Gd2O3 content in gadolinium fuel blends and gadolinium fuel pellets by atomic emission spectrometry using an inductively coupled plasma source (ICP-AES) |
This document is applicable to the determination of gadolinium as Gd2O3 in powder blends and sintered pellets of Gd2O3 + UO2 and ((U, Gd) O2) from mass fraction 10 g/kg to 100 g/kg (i.e. 1 % to 10 %), using a suitable ICP-AES instrument.
This methodology is capable of demonstrating compliance with agreed upon fuel specifications and associated data quality objectives provided the user has performed qualification measurements under their established measurement control program to demonstrate that measurement uncertainty requirements will be met with the desired level of confidence at the specification.
|
Published |
2022-01 |
Edition : 2 |
Number of pages : 9 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 16966:2013 |
Nuclear energy — Nuclear fuel technology — Theoretical activation calculation method to evaluate the radioactivity of activated waste generated at nuclear reactors |
ISO 16966:2013 gives guidelines for a common basic theoretical methodology to evaluate the activity of radionuclides in activated waste generated at nuclear reactors using neutron activation calculations.
|
Published |
2013-12 |
Edition : 1 |
Number of pages : 45 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-1:2007 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 1: Procedural overview |
ISO 18213-1:2007 describes procedures for tank calibration and volume determination for nuclear process tanks equipped with pressure-measurement systems for determining liquid content. Specifically, overall guidance is provided for planning a calibration exercise undertaken to obtain the data required for the equation to estimate a tank's volume and the key steps in the procedure are presented for subsequently using the estimated volume-measurement equation to determine tank liquid volumes.
The procedures presented apply specifically to tanks equipped with bubbler probe systems for measuring liquid content. Moreover, these procedures produce reliable results only for clear (i.e., without suspended solids), homogeneous liquids that are at both thermal and static equilibrium.
|
Published |
2007-11 |
Edition : 1 |
Number of pages : 22 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-2:2007 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 2: Data standardization for tank calibration |
ISO 18213-2:2007 presents procedures for standardizing a set of calibration data to a fixed set of reference conditions so as to minimize the effect of variations in ambient conditions that occur during the measurement process. The procedures presented herein apply generally to measurements of liquid height and volume obtained for the purpose of calibrating a tank (i.e., calibrating a tank's measurement system). When used in connection with other parts of ISO 18213, these procedures apply specifically to tanks equipped with bubbler probe systems for measuring liquid content.
The standardization algorithms presented in ISO 18213-2:2007 can be profitably applied when only estimates of ambient conditions, such as temperature, are available. However, the most reliable results are obtained when relevant ambient conditions are measured for each measurement of volume and liquid height in a set of calibration data.
|
Published |
2007-11 |
Edition : 1 |
Number of pages : 15 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-3:2009 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 3: Statistical methods |
ISO 18213-3:2009 presents statistical procedures that can be applied to tank calibration and volume measurement data for nuclear materials accountancy tanks. In particular, ISO 18213-3:2009 presents several diagnostic plots that can be used to evaluate and compare tank calibration data; a procedure for estimating the uncertainties of tank calibration measurements (i.e., determinations of height and volume); a model for estimating either a tank's calibration equation or its inverse (the measurement equation), together with related uncertainties, from a set of standardized tank calibration data (i.e., from a series of standardized height-volume determinations); and a method for computing uncertainty estimates for determinations of liquid volume.
It is intended that the methods in ISO 18213-3:2009 be used within the context of the other parts of ISO 18213. Specifically, the methods presented in ISO 18213-3:2009 are tailored to the general methodology described in ISO 18213-1 and to appropriate related algorithms in ISO 18213-2, ISO 18213-4, ISO 18213-5 or ISO 18213-6. Although the methodology in ISO 18213-3:2009 is intended for application specifically within the context of the other parts of ISO 18213, the methods are more widely applicable. In particular, the statistical model presented in Clause 5 for estimating the tank's measurement equation from a set of standardized calibration data can be applied, regardless of whether or not these data are acquired in accordance with the methods of ISO 18213. A similar statement holds for (propagation) methods of variance estimation: it is intended that the results in ISO 18213-3:2009 be applied to the specific models for which they were derived, but the methods themselves are more widely applicable.
An option is presented for a facility to develop equivalent plant- or tank-specific methods of statistical analysis as an alternative to ISO 18213-3:2009. However, if a facility adopts ISO 18213 and chooses not to develop equivalent alternative methods of statistical analysis, it is necessary to use the methods of ISO 18213-3:2009.
|
Published |
2009-03 |
Edition : 1 |
Number of pages : 49 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-4:2008 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 4: Accurate determination of liquid height in accountancy tanks equipped with dip tubes, slow bubbling rate |
ISO 18213-4:2008 specifies a procedure for making accurate determinations of the liquid height in nuclear-materials-accountancy tanks that are equipped with pneumatic systems for determining the liquid content. With such systems, gas is forced through a probe (dip tube) whose tip is submerged in the tank liquid. The pressure required to induce bubbling is measured with a manometer located at some distance from the tip of the probe. This procedure applies specifically when a very slow bubbling rate is employed.
|
Published |
2008-03 |
Edition : 1 |
Number of pages : 21 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-5:2008 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 5: Accurate determination of liquid height in accountancy tanks equipped with dip tubes, fast bubbling rate |
ISO 18213-5:2008 specifies a procedure for making accurate determinations of liquid height in nuclear-materials-accountancy tanks that are equipped with pneumatic systems for determining the liquid content. With such systems, gas is forced through a probe (dip tube) whose tip is submerged in the tank liquid. The pressure required to induce bubbling is measured with a manometer located at some distance from the tip of the probe. This procedure applies specifically when a fast bubbling rate is employed.
|
Published |
2008-03 |
Edition : 1 |
Number of pages : 13 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18213-6:2008 |
Nuclear fuel technology — Tank calibration and volume determination for nuclear materials accountancy — Part 6: Accurate in-tank determination of liquid density in accountancy tanks equipped with dip tubes |
ISO 18213-6:2008 specifies a procedure for making accurate determinations of the densities of process liquids from in-tank measurements of the liquid content. This procedure is applicable to tanks equipped with pneumatic systems for determining the liquid content that have two or more bubbler probes of differing lengths. The probes must be fixed relative to each other and to the tank in which they are installed.
|
Published |
2008-03 |
Edition : 1 |
Number of pages : 12 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18256-1:2019 |
Nuclear fuel technology — Dissolution of plutonium dioxide-containing materials — Part 1: Dissolution of plutonium dioxide powders |
This document specifies the dissolution of powder samples of plutonium oxide for subsequent determination of elemental concentration and isotopic composition.
|
Published |
2019-01 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18256-2:2019 |
Nuclear fuel technology — Dissolution of plutonium dioxide-containing materials — Part 2: Dissolution of MOX pellets and powders |
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.
|
Published |
2019-01 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18315:2018 |
Nuclear energy — Guidance to the evaluation of measurement uncertainties of impurity in uranium solution by linear regression analysis |
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".
|
Published |
2018-11 |
Edition : 1 |
Number of pages : 18 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 18557:2017 |
Characterisation principles for soils, buildings and infrastructures contaminated by radionuclides for remediation purposes |
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.
|
Published |
2017-09 |
Edition : 1 |
Number of pages : 38 |
Technical Committee |
13.020.40
Pollution, pollution control and conservation
;
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21847-2:2007 |
Nuclear fuel technology — Alpha spectrometry — Part 2: Determination of plutonium in uranium and its compounds |
ISO 21847-2:2007 describes a method for determining trace amounts of 238Pu, 239Pu + 240Pu in uranium hexafluoride, uranium oxides or uranyl nitrate.
|
Published |
2007-09 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 19017:2015 |
Guidance for gamma spectrometry measurement of radioactive waste |
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).
|
Published |
2015-12 |
Edition : 1 |
Number of pages : 48 |
Technical Committee |
17.240
Radiation measurements
;
27.120.30
Fissile materials and nuclear fuel technology
|
| 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 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.
|
Published |
2007-04 |
Edition : 1 |
Number of pages : 23 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21391:2019 |
Nuclear criticality safety — Geometrical dimensions for subcriticality control — Equipment and layout |
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.
|
Published |
2019-08 |
Edition : 1 |
Number of pages : 14 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21483:2013 |
Determination of solubility in nitric acid of plutonium in unirradiated mixed oxide fuel pellets (U, Pu) O2 |
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.
|
Published |
2013-11 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21484:2008 |
Nuclear fuel technology — Determination of the O/M ratio in MOX pellets — Gravimetric method |
ISO 21484:2008 describes a method for determining the oxygen-to-metal (O/M) ratio in mixed uranium-plutonium oxide (U,Pu)O2 ± X pellets.
|
Withdrawn |
2008-11 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
17.240
Radiation measurements
;
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 21484:2017 |
Nuclear Energy — Fuel technology — Determination of the O/M ratio in MOX pellets by the gravimetric method |
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.
|
Published |
2017-01 |
Edition : 2 |
Number of pages : 6 |
Technical Committee |
17.240
Radiation measurements
;
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 6676:1990 |
Acid-grade and ceramic-grade fluorspar — Determination of total phosphorus content — Reduced-molybdophosphate spectrometric method |
|
Withdrawn |
1990-03 |
Edition : 2 |
Number of pages : 3 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 22765:2016 |
Nuclear fuel technology — Sintered (U,Pu)O2 pellets — Guidance for ceramographic preparation for microstructure examination |
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.
|
Published |
2016-12 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 22875:2008 |
Nuclear energy — Determination of chlorine and fluorine in uranium dioxide powder and sintered pellets |
ISO 22875:2008 describes a method for determining chlorine and fluorine in uranium dioxide powder and sintered pellets. It is applicable for the analysis of samples with from 5 µg to 500 µg of chlorine per gram sample and with 2 µg to 500 µg of fluorine per gram sample.
|
Withdrawn |
2008-01 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 22875:2017 |
Nuclear energy — Determination of chlorine and fluorine in uranium dioxide powder and sintered pellets |
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.
|
Published |
2017-08 |
Edition : 2 |
Number of pages : 12 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 22946:2020 |
Nuclear criticality safety — Solid waste excluding irradiated and non-irradiated nuclear fuel |
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.
|
Published |
2020-01 |
Edition : 1 |
Number of pages : 11 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 24459:2021 |
Determination of uranium content in samples coming from the nuclear fuel cycle by L-absorption edge spectrometry |
This document specifies a method for the determination of uranium concentrations in nitric acid or TBP-DILUANT (for example TBP-kerosene) solutions coming from the nuclear fuel cycle.
The method is applicable
— for process control of solutions, free of suspension, which contain between 10 g/l to 300 g/l uranium, and
— for high accuracy purposes (Safeguards) to nitric acid solutions, free of suspension, which contain between 100 g/l and 220 g/l uranium.
Having
— the content of neptunium and plutonium impurities in the solution less than 1 % of the uranium content.
— the content of neutron poisons (gadolinium, erbium) less than 1 % of the uranium content to ensure the absence of significant interferences at the level of required precision, for high accuracy purposes.
The method is applicable to solid samples as well, provided that they can be fully dissolved in nitric acid.
|
Published |
2021-10 |
Edition : 1 |
Number of pages : 13 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 26062:2010 |
Nuclear technology — Nuclear fuels — Procedures for the measurement of elemental impurities in uranium- and plutonium-based materials by inductively coupled plasma mass spectrometry |
ISO 26062:2010 specifies a procedure for the determination of trace impurities in uranium- or plutonium‑based, or mixed uranium- and plutonium‑based, materials by inductively coupled plasma mass spectrometry (ICP-MS). It provides both guidelines and specific options for the determination of an element or group of elements.
It is applicable to solutions such as uranyl or plutonium nitrate, solids such as the oxides and to mixed actinide materials such as unirradiated mixed oxide material in either solid or dissolved forms. It is not directly suitable for the analysis of uranium or plutonium matrices containing significant quantities of other elements such as uranium–gadolinium mixtures. It may nevertheless form the basis of a process for analysing this type of matrix, provided that the impact of the gadolinium component is ascertained.
|
Published |
2010-09 |
Edition : 1 |
Number of pages : 21 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO/IEC 80079-20-2:2016/Cor 1:2017 |
Explosive atmospheres — Part 20-2: Material characteristics — Combustible dusts test methods — Technical Corrigendum 1 |
|
Published |
2017-03 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
29.260.20
Electrical apparatus for explosive atmospheres
|
| ISO 27467:2009 |
Nuclear criticality safety — Analysis of a postulated criticality accident |
ISO 27467:2008 specifies areas that it is important to study when analyzing potential criticality accidents.
It is important that a criticality accident analysis be performed each time a criticality accident is considered credible, due either to criticality contingencies (double batching, procedural violations, etc.) or to the failure of safety provisions (effectiveness of neutron absorber reduced by fire, etc.).
It is important that the criticality safety specialist be mindful that the process of evaluation developed in ISO 27467:2008 does not cater to the unforeseen, since any actually occurring criticality accident will probably result from a scenario not envisioned or from failure to comply with prevailing regulations.
ISO 27467:2008 does not address detailed administrative measures, for which the responsibility lies with the public authorities, nor does it deal with criteria used to justify the accident criticality analysis of a nuclear facility.
ISO 27467:2008 does not apply to nuclear power plants.
|
Published |
2009-02 |
Edition : 1 |
Number of pages : 7 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 27468:2011 |
Nuclear criticality safety — Evaluation of systems containing PWR UOX fuels — Bounding burnup credit approach |
ISO 27468:2011 establishes an evaluation methodology for nuclear criticality safety with burnup credit. It identifies important parameters and specifies requirements, recommendations, and precautions to be taken into account in the evaluations. It also highlights the main important technical fields to ensure that the fuel composition or history considered in calculations provides a bounding value of the effective neutron multiplication factor, keff.
ISO 27468:2011 is applicable to transport, storage, disposal or reprocessing units implying irradiated fissile material from pressurized water reactor (PWR) fuels that initially contain uranium oxide (UOX).
Fuels irradiated in other reactors (e.g. boiling water reactors) and fuels that initially contain mixed uranium-plutonium oxide are not covered in ISO 27468:2011.
ISO 27468:2011 does not specify requirements related to overall criticality safety evaluation or eventual implementation of burnup credit.
|
Published |
2011-07 |
Edition : 1 |
Number of pages : 10 |
Technical Committee |
27.120.30
Fissile materials and nuclear fuel technology
|
| ISO 67:1981 |
Muscovite mica blocks, thins and films — Grading by size |
|
Withdrawn |
1981-11 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 444:1981 |
Phlogopite mica blocks, thins and splittings — Grading by size |
|
Withdrawn |
1981-11 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 2185:1972 |
Muscovite mica blocks, thins and films — Visual classification |
|
Withdrawn |
1972-11 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 3703:1976 |
Acid-grade fluorspar — Determination of flotation agents |
|
Withdrawn |
1976-05 |
Edition : 1 |
Number of pages : 2 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 3703:1993 |
Acid-grade and ceramic-grade fluorspar — Determination of flotation agents |
The principle of the method is treatment of a test portion with a mixture of dilute hydrochloric acid and an organic solvent, removal of the insoluble fluorspar by filtration under vacuum, separation of the organic phase containing the flotation agent, evaporation of the solvent and weighing of the residue. Applies to materials which have been subjected to flotation process and which have flotation agent contents equal to or greater than 0,002 % (m/m) of the dried material.
|
Withdrawn |
1993-03 |
Edition : 2 |
Number of pages : 2 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4282:1977 |
Acid-grade fluorspar — Determination of loss in mass at 105 degrees C |
|
Withdrawn |
1977-02 |
Edition : 1 |
Number of pages : 2 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4282:1992 |
Acid-grade and ceramic-grade fluorspar — Determination of loss in mass at 105 degrees C |
The method is based on drying a test portion at 105 °C and determination of the loss in mass, which corresponds to the content of water and other components volatile at that temperature. Applies to fluorspars, which may be dried material containing not less than 0,02 % (m/m) of components volatile at 105 °C, or filter cake.
|
Withdrawn |
1992-12 |
Edition : 2 |
Number of pages : 2 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4283:1978 |
Acid-grade fluorspar — Determination of carbonate content — Titrimetric method |
|
Withdrawn |
1978-09 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4283:1990 |
All grades of fluorspar — Determination of carbonate content — Titrimetric method |
|
Withdrawn |
1990-06 |
Edition : 2 |
Number of pages : 5 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 9505:1992 |
All grades of fluorspar — Determination of arsenic content — Silver diethyldithiocarbamate spectrometric method |
Specifies the principle, the reagents, the apparatus, the preparation of the test sample, the test procedure, the expression of results and the test report.
|
Withdrawn |
1992-01 |
Edition : 1 |
Number of pages : 6 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4283:1993 |
All grades of fluorspar — Determination of carbonate content — Titrimetric method |
The principle of the method specified is treatment of a test portion with hydrochloric acid solution, absorption of the evolved carbon dioxide in barium hydroxide solution, neutralization of excess alkali with hydrochloric acid solution, addition of an exactly measured excess of a standard volumetric hydrochloric acid solution to dissolve the precipitated barium carbonate and back-titration with a standard volumetric sodium hydroxide solution using methyl orange as indicator. The method is applicable to products having carbonate contents equal to or greater than 0,04 % (m/m).
|
Withdrawn |
1993-01 |
Edition : 3 |
Number of pages : 5 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4284:1978 |
Acid-grade fluorspar — Determination of sulphide content — Iodometric method |
|
Withdrawn |
1978-07 |
Edition : 1 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4284:1988 |
Acid-grade and ceramic-grade fluorspar — Determination of sulfide content — Iodometric method |
|
Withdrawn |
1988-11 |
Edition : 2 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 4284:1993 |
Acid-grade and ceramic-grade fluorspar — Determination of sulfide content — Iodometric method |
The principle of the method specified is digestion of a test portion in a sealed apparatus in a mixture of hydrochloric acid, tin(II) chloride and boric acid solutions, absorption of the liberated hydrogen sulfide, entrained in a stream of oxygen-free argon or nitrogen, in zinc acetate solution and iodometric determination of the zinc sulfide formed. The method is applicable to products having a sulfide content equal to or greater than 0,001 % (m/m). It is not applicable, if presence of polysulfides is suspected.
|
Withdrawn |
1993-01 |
Edition : 3 |
Number of pages : 4 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 5023:1977 |
Phlogopite mica splittings — Thermal classification |
|
Withdrawn |
1977-12 |
Edition : 1 |
Number of pages : 1 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 5437:1988 |
Acid-grade and ceramic-grade fluorspar — Determination of barium sulfate content — Gravimetric method |
|
Withdrawn |
1988-11 |
Edition : 1 |
Number of pages : 2 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 5437:1992 |
Acid-grade and ceramic-grade fluorspar — Determination of barium sulfate content — Gravimetric method |
The method is based on evaporation to dryness of a test portion in the presence of hydrofluoric acid and concentrated sulfuric acid, extraction of the soluble salts from the mixture, filtration, drying and weighing of the residual barium sulfate. Applies to products having barium sulfate contents equal to or greater than 0,1 % (m/m).
|
Withdrawn |
1992-12 |
Edition : 2 |
Number of pages : 3 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 5438:1985 |
Acid-grade and ceramic-grade fluorspar — Determination of silica content — Reduced molybdosilicate spectrometric method |
|
Withdrawn |
1985-06 |
Edition : 1 |
Number of pages : 5 |
Technical Committee |
73.080
Non-metalliferous minerals
|
| ISO 5438:1993 |
Acid-grade and ceramic-grade fluorspar — Determination of silica content — Reduced-molybdosilicate spectrometric method |
The method specified is based on decomposition of a test portion by fusion with sodium carbonate and subsequent acidification with hydrochloric acid in the presence of boric acid to complex fluoride, formation of the molybdosilicic acid complex and selective reduction to the blue molybdosilicic acid complex with addition of tartaric acid to prevent interference from phosphate, spectrometric measurement of the absorbance of the coloured complex at the wavelength of maximum absorption (about 795 nm). Applies to products having silica contents, expressed as SiO2, in the range 0,05 % (m/m) to 4,0 % (m/m).
|
Withdrawn |
1993-03 |
Edition : 2 |
Number of pages : 5 |
Technical Committee |
73.080
Non-metalliferous minerals
|