CEN/TC 430 - Nuclear energy, nuclear technologies, and radiological protection

This document specifies a screening test method to quantify rapidly the activity concentration of gamma-emitting radionuclides, such as 131I, 132Te, 134Cs and 137Cs, in solid or liquid test samples using gamma-ray spectrometry with lower resolution scintillation detectors as compared with the HPGe detectors (see IEC 61563[7]).
This test method can be used for the measurement of any potentially contaminated environmental matrices (including soil), food and feed samples as well as industrial materials or products that have been properly conditioned[8]. Sample preparation techniques used in the screening method are not specified in this document, since special sample preparation techniques other than simple machining (cutting, grinding, etc.) should not be required. Although the sampling procedure is of utmost importance in the case of the measurement of radioactivity in samples, it is out of scope of this document; other International Standards for sampling procedures that can be used in combination with this document are available (see References [9] [10] [11] [12] [13] [14]).
The test method applies to the measurement of gamma-emitting radionuclides such as 131I, 134Cs and 137Cs. Using sample sizes of 0,5 l to 1,0 l in a Marinelli beaker and a counting time of 5 min to 20 min, decision threshold of 10 Bq·kg−1 can be achievable using a commercially available scintillation spectrometer [e.g. thallium activated sodium iodide (NaI(Tl)) spectrometer 2” ϕ × 2” (50,8 mm Ø x 50,8 mm) detector size, 7 % resolution (FWHM) at 662 keV, 30 mm lead shield thickness].
This test method also can be performed in a “makeshift” laboratory or even outside a testing laboratory on samples directly measured in the field where they were collected.
During a nuclear or radiological emergency, this test method enables a rapid measurement of the activity concentration of potentially contaminated samples to check against operational intervention levels (OILs) set up by decision makers that would trigger a predetermined emergency response to reduce existing radiation risks[2].
Due to the uncertainty associated with the results obtained with this test method, test samples requiring more accurate test results can be measured using high purity germanium (HPGe) detectors gamma-ray spectrometry in a testing laboratory, following appropriate preparation of the test samples[15][16].
This document does not contain criteria to establish the activity concentration of OILs.

The purpose of this document is to give an overview of the minimum requirements for performing the dicentric assay with quality control measures using mitogen stimulated peripheral blood lymphocytes for initial assessment of individuals involved in a mass casualty scenario. The dicentric assay is the use of chromosome damage to quickly estimate approximate radiation doses received by individuals in order to supplement the early clinical categorization of casualties.
This document focuses on the organizational and operational aspects of applying the dicentric assay in an initial assessment mode. The technical aspects of the dicentric assay can be found in ISO 19238.
This document is applicable either to an experienced biological dosimetry laboratory working alone or to a network of collaborating laboratories (as defined in Clause 7).

This document describes a generic test method for measuring alpha emitting radionuclides, for all types of samples (soil, sediment, construction material, foodstuff, water, airborne, environmental bio-indicator, human biological samples as urine, faeces etc.) by alpha spectrometry. This method can be used for any type of environmental study or monitoring of alpha emitting radionuclides activities.
If relevant, this test method requires appropriate sample pre-treatment followed by specific chemical separation of the test portion in order to obtain a thin source proper to alpha spectrometry measurement.
This test method can be used to determine the activity, specific activity or activity concentration of a sample containing alpha emitting radionuclides such as 210Po, 226Ra, 228Th, 229Th, 230Th, 232Th, 232U,234U, 235U, 238U, 238Pu, 239+240Pu, 241Am or 243+244Cm.
This test method can be used to measure very low levels of activity, one or two orders of magnitude less than the usual natural levels of alpha emitting radionuclides. Annexes B of UNSCEAR 2000 and UNSCEAR 2008  give, respectively, typical natural activity concentrations for air, foods, drinking waters and, soils and building materials. The detection limit of the test method depends on the amount of the sample material analysed (mass or volume) after concentration, chemical yield, thickness of measurement source and counting time.
The quantity of the sample to be collected and analysed depends on the expected activity of the sample and the detection limit to achieve.

This document specifies methods and means of monitoring for inadvertent movement and illicit trafficking of radioactive material. It provides guidelines on the use of both stationary and portable, for example hand-held, instruments to monitor for radiation signatures from radioactive material. Emphasis is placed on the operational aspects, i.e., requirements derived for monitoring of traffic and commodities mainly at border-crossing facilities. Although the term border is used repeatedly in this document, it is meant to apply not only to international land borders but also maritime ports, airports, and similar locations where goods or individuals are being checked. This document does not specifically address the issue of detection of radioactive materials at recycling facilities, although it is recognized that transboundary movement of metals for recycling occurs, and that monitoring of scrap metals might be done at the borders of a state.
This document is applicable to
—    regulatory bodies and other competent authorities seeking guidance on implementation of action plans to combat illicit trafficking,
—    law enforcement agencies, for example border guards, to obtain guidelines on recommended monitoring procedures,
—    equipment manufacturers in order to understand minimum requirements derived from operational necessities according to this document, and
—    end-users of radiation detection equipment applicable to this document.

This document describes procedures for calibrating and determining the response of dosemeters and dose-rate meters in terms of the operational quantities for radiation protection purposes defined by the International Commission on Radiation Units and Measurements (ICRU). However, as noted in ICRU 56, the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is a guide for those who calibrate protection-level dosemeters and dose-rate meters with beta-reference radiation and determine their response as a function of beta-particle energy and angle of incidence. Such measurements can represent part of a type test during the course of which the effect of other influence quantities on the response is examined. This document does not cover the in-situ calibration of fixed, installed area dosemeters. The term “dosemeter” is used as a generic term denoting any dose or dose-rate meter for individual or area monitoring. In addition to the description of calibration procedures, this document includes recommendations for appropriate phantoms and the way to determine appropriate conversion coefficients. Guidance is provided on the statement of measurement uncertainties and the preparation of calibration records and certificates.

This document specifies the applicable requirements related to the design and the operation of confinement and ventilation systems for fusion facilities for tritium fuels and tritium fuel handling facilities specific for fusion applications for peaceful purposes using high tritium inventories, as well as for their specialized buildings such as hot cells, examination laboratories, emergency management centres, radioactive waste treatment and storage facilities.
In most countries, a tritium quantity is declared as high for tritium inventories higher than a range of 10 g to 100 g. In the tritium fusion facilities in the scope of this document, the tritium inventory is deemed to be higher than this range for the whole site.
This document applies especially to confinement and ventilation systems that ensure the safety function of nuclear facilities involved in nuclear fusion with the goal to protect the workers, the public and the environment from the dissemination of radioactive contamination originating from the operation of these installations, and in particular from airborne tritium contamination with adequate confinement systems.

This document specifies the requirements for reference beta radiation fields produced by radioactive sources to be used for the calibration of personal and area dosemeters and dose-rate meters to be used for the determination of the quantities Hp(0,07), H'(0,07;Ω), Hp(3) and H'(3;Ω), and for the determination of their response as a function of beta particle energy and angle of incidence. The basic quantity in beta dosimetry is the absorbed-dose rate in a tissue-equivalent slab phantom. This document gives the characteristics of radionuclides that have been used to produce reference beta radiation fields, gives examples of suitable source constructions and describes methods for the measurement of the residual maximum beta particle energy and the dose equivalent rate at a depth of 0,07 mm in the International Commission on Radiation Units and Measurements (ICRU) sphere. The energy range involved lies between 0,22 MeV and 3,6 MeV maximum beta energy corresponding to 0,07 MeV to 1,2 MeV mean beta energy and the dose equivalent rates are in the range from about 10 µSv·h-1 to at least 10 Sv·h-1.. In addition, for some sources, variations of the dose equivalent rate as a function of the angle of incidence are given. However, as noted in ICRU 56[5], the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is applicable to two series of reference beta radiation fields, from which the radiation necessary for determining the characteristics (calibration and energy and angular dependence of response) of an instrument can be selected.
Series 1 reference radiation fields are produced by radioactive sources used with beam-flattening filters designed to give uniform dose equivalent rates over a large area at a specified distance. The proposed sources of 106Ru/106Rh, 90Sr/90Y, 85Kr, 204Tl and 147Pm produce maximum dose equivalent rates of approximately 200 mSv·h–1.
Series 2 reference radiation fields are produced without the use of beam-flattening filters, which allows large area planar sources and a range of source-to-calibration plane distances to be used. Close to the sources, only relatively small areas of uniform dose rate are produced, but this series has the advantage of extending the energy and dose rate ranges beyond those of series 1. The series also include radiation fields using polymethylmethacrylate (PMMA) absorbers to reduce the maximum beta particle energy. The radionuclides used are those of series 1; these sources produce dose equivalent rates of up to 10 Sv·h–1.

This document provides requirements and guidance regarding the use of CAAS for operations of a nuclear facility. Requirements and guidance on CAAS design are provided in the IEC 60860.
This document is applicable to operations with fissile materials outside nuclear reactors but within the boundaries of nuclear establishments.
This document applies when a need for CAAS has been established. Information about the need for CAAS is given in Annex C.
This document does not include details of administrative steps, which are considered to be activities of a robust management system (ISO 14943 provides details of administrative steps).
Details of nuclear accident dosimetry and personnel exposure evaluations are not within the scope of this document.
This document is concerned with gamma and neutron radiation rate-sensing systems. Specific detection criteria can also be met with integrating systems; systems detecting either neutron or gamma radiation can also be used. Equivalent considerations then apply.

This document specifies methods for the measurement of the absorbed-dose rate in a tissue-equivalent slab phantom in the ISO 6980 reference beta-particle radiation fields. The energy range of the beta-particle-emitting isotopes covered by these reference radiations is 0,22 MeV to 3,6 MeV maximum beta energy corresponding to 0,07 MeV to 1,2 MeV mean beta energy. Radiation energies outside this range are beyond the scope of this document. While measurements in a reference geometry (depth of 0,07 mm or 3 mm at perpendicular incidence in a tissue‑equivalent slab phantom) with an extrapolation chamber used as primary standard are dealt with in detail, the use of other measurement systems and measurements in other geometries are also described, although in less detail. However, as noted in ICRU 56, the ambient dose equivalent, H*(10), used for area monitoring, and the personal dose equivalent, Hp(10), as used for individual monitoring, of strongly penetrating radiation, are not appropriate quantities for any beta radiation, even that which penetrates 10 mm of tissue (Emax > 2 MeV).
This document is intended for those organizations wishing to establish primary dosimetry capabilities for beta particles and serves as a guide to the performance of dosimetry with an extrapolation chamber used as primary standard for beta‑particle dosimetry in other fields. Guidance is also provided on the statement of measurement uncertainties.

This document provides procedures for monitoring the dose to the skin, the extremities, and the lens of the eye. It gives guidance on how to decide if such dosemeters are needed and to ensure that individual monitoring is appropriate to the nature of the exposure, taking practical considerations into account.
This document specifies procedures for individual monitoring of radiation exposure of the skin of the body, extremities (skin of the hands, fingers, wrists, forearms including elbow, lower leg including patella, feet and ankles), and lens of the eye in planned exposure situations. It covers practices which involve a risk of exposure to photons in the range of 8 keV to 10 MeV, electrons and positrons in the range of 0,07 MeV to 1,2 MeV mean beta energies being equivalent to 0,22 MeV and 3,6 MeV beta maximum energy - in accordance to the ISO 6980 series, and neutrons in the range of thermal to 20 MeV.
This document gives guidance for the design of a monitoring programme to ensure compliance with legal individual dose limits. It refers to the appropriate operational dose quantities, and it gives guidance on the type and frequency of individual monitoring and the type and positioning of the dosemeter. Finally, different approaches to assess and analyse skin, extremity, and lens of the eye doses are given.
It is not in the scope of this document to consider exposure due to alpha radiation fields.
NOTE 1        The requirements for the monitoring of the occupational exposure may be given in national regulations.
NOTE 2        Dose to the lens of the eye due to intake of tritium is not in the scope of this document. Moreover, the situation of the workers that work in contaminated atmosphere and can have alpha and/or radon eye lens dose is also not in the scope.

This document specifies a method which covers the determination of Gd2O3 content in UO2 fuel pellets, by X-ray fluorescence spectrometry.
Either wave dispersion X-ray fluorescence (WD-XRF) or energy dispersion X-ray fluorescence (ED-XRF) is applicable, however, this document states a method by using WD-XRF using Gd Lα-line.
This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.

This document applies to the determination of beta emitters activity concentration using liquid scintillation counting. The method requires the preparation of a scintillation source, which is obtained by mixing the test sample and a scintillation cocktail. The test sample can be liquid (aqueous or organic), or solid (particles or filter or planchet).
NOTE            Planchet are samples, described in 8.5, out of solid material e.g. small metal, plastic or glass pans or support material made of these materials
This document describes the conditions for measuring the activity concentration of beta emitter radionuclides by liquid scintillation counting[2].
The choice of the test method using liquid scintillation counting involves the consideration of the potential presence of other beta-, alpha- and gamma emitter radionuclides in the test sample. In this case, a specific sample treatment by separation or extraction is implemented to isolate the radionuclide of interest in order to avoid any interference with other beta-, alpha- and gamma-emitting radionuclides during the counting phase.
This document is applicable to all types of liquid samples having an activity concentration ranging from about 1 Bq·l−1 to 106 Bq·l−1. For a liquid test sample, it is possible to dilute liquid test samples in order to obtain a solution having an activity compatible with the measuring instrument. For solid samples, the activity of the prepared scintillation source shall be compatible with the measuring instrument.
The measurement range is related to the test method used: nature of test portion, preparation of the scintillator - test portion mixture, measuring assembly as well as to the presence of the co-existing activities due to interfering radionuclides.
Test portion preparations (such as distillation for 3H measurement, or benzene synthesis for 14C measurement, etc.) are outside the scope of this document and are described in specific test methods using liquid scintillation[3][[4][5][6][7][8][9][10].

This document is applied to fuel fabrication. It describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the (U,Pu)O2 pellet microstructure.
The examinations are performed
a)       before any treatment or any etching, and
b)       after thermal treatment or after chemical or ion 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 or by alpha autoradiography. A scanning electron microscope (SEM) or a microprobe can also be used. In this case an additional preparation can be needed depending on the equipment used. This preparation is not in the scope of this standard.

This document specifies the minimum requirements for the design of programmes to monitor workers exposed to the risk of internal contamination by radioactive material and establishes principles for the development of compatible goals and requirements for monitoring programmes.
This document specifies the
a)       purposes of monitoring and monitoring programmes,
b)       description of the different categories of monitoring programmes,
c)        quantitative criteria for conducting monitoring programmes,
d)       suitable monitoring methods and criteria for their selection,
e)       information that has to be collected for the design of a monitoring programme,
f)         general requirements for monitoring programmes (e.g. detection limits, tolerated uncertainties),
g)       frequencies of measurements calculated using the ICRP Occupational Intakes of Radionuclides (OIR) series,
h)       individual monitoring in specific cases (intake of actinides, intake via a wound and intake through the intact skin),
i)         quality assurance, and
j)         documentation, reporting and record-keeping.
This document does not apply to
—     the monitoring of exposure to radon and its radioactive decay products,
—     detailed descriptions of measuring methods and techniques,
—     detailed procedures for in vivo measurements and in vitro analysis,
—     interpretation of measurements results in terms of dose,
—     biokinetic data and mathematical models for converting measured activities into absorbed dose, equivalent dose and effective dose,
—     the investigation of the causes or implications of an exposure or intake.

This document provides guidance for
—   the sampling process of the aerosol particles in the air using filter media. This document takes into account the specific behaviour of aerosol particles in ambient air.
—   Two methods for sampling procedures with subsequent or simultaneous measurement:
—   the determination of the activity concentration of radionuclides bound to aerosol particles in the air knowing the activity deposited in the filter;
—   the operating use of continuous air monitoring devices used for real time measurement.
This document describes the test method to determine activity concentrations of radionuclides bound to aerosol particles after air sampling passing through a filter media designed to trap aerosol particles. The method can be used for any type of environmental study or monitoring.
This document does not cover the details of measurement test techniques (gamma spectroscopy, global alpha and beta counting, liquid scintillation, alpha spectrometry) used to determine the activity deposited in the media filter, which are either based on existing standards or internal methods developed by the laboratory in charge of those measurements. Also, this document does not cover the variability of the aerosol particle sizes as given by the composition of the dust contained in ambient air. This document does not address to sampling of radionuclides bound to aerosol particles in the effluent air of nuclear facilities [see ISO 2889:2021].

This document specifies general requirements for proficiency tests that are offered to in vivo bioassay measurement facilities operating a whole-body counter (WBC) or partial body counter (PBC) for monitoring of persons.
This document covers proficiency tests that involve only the quantification of radionuclides and tests that require the identification of radionuclides and their activity.
This document does not define specific requirements on administrative aspects of proficiency testing, such as shipping and finance, that may be the subject of national or international regulation.

This document specifies the dosimetric and organizational criteria and the test procedures to be used for the periodic verification of the performance of dosimetry services supplying personal and/or area, i.e. workplace and/or environmental, dosemeters used for individual (personal) and/or area, i.e. workplace and/or environmental monitoring.
NOTE            The quality of a supplier of a dosimetry service depends on both the characteristics of the approved (type‑tested) dosimetry system and the training and experience of the staff, together with the calibration procedures and quality assurance programmes.
The performance evaluation according to this document can be carried out by a dosimetry service to demonstrate the fulfilment of specified performance requirements. The irradiation qualities used in this document are representative for exposure situations that are expected or mimic workplace fields from the radiological activities being monitored using the dosemeters from the services.
This document applies to personal and area dosemeters for the assessment of external photon radiation with a fluence-weighted mean energy between 8 keV and 10 MeV, beta radiation with a fluence-weighted mean energy between 60 keV and 1,2 MeV, and neutron radiation with a fluence-weighted mean energy between 25,3 meV, i.e. thermal neutrons with a Maxwellian energy distribution with kT = 25,3 meV, and 200 MeV.
It covers all types of personal and area dosemeters needing laboratory processing (e.g. thermoluminescent, optically stimulated luminescence, radiophotoluminescent, track detectors or photographic-film dosemeters) and involving continuous measurements or measurements repeated regularly at fixed time intervals (e.g. several weeks, one month).
Active direct reading as well as semi-passive or hybrid dosemeters, such as direct ion storage (DIS) or silicon photomultiplier (SiPM) dosemeters, for dose measurement, can also be treated according to this document. Then, they are treated as if they were passive, i.e. the dosimetry service reads their indicated values and reports them to the evaluation organization.
In this document, the corrected indicated (corrected indication) value is the one given by the dosimetry systems as the final result of the evaluation algorithm (for example display of the software, printout) in units of dose equivalent (Sv).
Environmental dosemeters usually indicate the quantity H*(10) but they can, in addition or alternatively, indicate the quantity H'(3), H'(0,07), air kerma, Ka, or absorbed dose, D. All these dosemeters can also be treated according to this document. If Ka or D is indicated (in Gy) the dose values in this document stated in Sv shall then be interpreted as equivalent values in Gy.

This document provides guidance for those who calibrate protection-level dosemeters and doserate meters for area and individual monitoring with reference neutron radiation fields. This includes the determination of the response as a function of neutron energy and angle of incidence. The operational quantities recommended in ICRU Report 51 are considered. In addition to the description of procedures, this document includes appropriate definitions and conversion coefficients and provides guidance on the statement of measurement uncertainties.

This document applies to the testing of the decontamination of textiles, which are contaminated by radioactive materials.
The test method describes the technique to assess the efficiency of decontamination agents (see ISO 7503-1 and ISO 7503-3).
This document applies to the testing of detergents, which may be used in aqueous solutions for the purpose of cleaning radioactively contaminated textiles.
The radionuclides used in this test are those commonly found in the nuclear industry (60Co and 137Cs or 134Cs) in aqueous form. The test can also be adapted for use with other radionuclides and other chemical forms, depending on the customer requirements, if the solutions are chemically stable and do not damage the test specimen.
The test method is not suitable if the radionuclide emits low energy gamma rays, like 55Fe, or low energy beta or alpha particles that are readily attenuated in the textile fabrics, or if the nuclide has a chemical or isotopic interaction with the detergent used in the method (e.g. tritium which could be in several chemical forms).
The test method does not apply to the testing of the ability of detergents to remove non-radioactive dirt.

This document describes a test method to determine the activity concentration of atmospheric tritium by trapping tritium in air by bubbling through a water solution.
The formulae are given for a sampling system with four bubblers. They can also be applied to trapping systems with only one trapping module consisting of two bubblers if only tritiated water vapour (HTO) is in the atmosphere to be sampled.
This document does not cover laboratory test sample results, in becquerel per litre of trapping solution, according to ISO 9698 or ISO 13168.
The test method detection limit result is between 0,2 Bq∙m-3 and 0,5 Bq∙m-3 when the sampling duration is about one week.

This document specifies the general requirements, based on ISO 11074 and ISO/IEC 17025, for all steps in the planning (desk study and area reconnaissance) of the sampling and the preparation of samples for testing. It includes the selection of the sampling strategy, the outline of the sampling plan, the presentation of general sampling methods and equipment, as well as the methodology of the pre-treatment of samples adapted to the measurements of the activity of radionuclides in soil including granular materials of mineral origin which contain NORM or artificial radionuclides, such as sludge, sediment, construction debris, solid waste of different type and materials from technologically enhanced naturally occurring radioactive materials (mining, coal combustion, phosphate fertilizer production etc.).

This document gives guidance on
a)       confidentiality of personal information for the customer and the laboratory,
b)       laboratory safety requirements,
c)        calibration sources and calibration dose ranges useful for establishing the reference dose-response curves that contribute to the dose estimation from CBMN assay yields and the detection limit,
d)       performance of blood collection, culturing, harvesting, and sample preparation for CBMN assay scoring,
e)       scoring criteria,
f)         conversion of micronucleus frequency in BNCs into an estimate of absorbed dose,
g)       reporting of results,
h)       quality assurance and quality control, and
i)         informative annexes containing sample instructions for customers, sample questionnaire, a microscope scoring data sheet, and a sample report.
This document excludes methods for automated scoring of CBMN.

This document specifies an analytical method for the electrochemical measurement of pure plutonium nitrate solutions of nuclear grade, with an expanded uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, k = 2). The method is applicable for aqueous solutions containing plutonium at more than 0,5 g/l and test samples containing plutonium between 4 mg and 15 mg. Application of this technique to solutions containing plutonium at less than 0,5 g/l and test samples containing plutonium at less than 4 mg requires experimental demonstration by the user that applicable data quality objectives will be met.

This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443.
NOTE          This document can be used as a criteria document for accreditation, peer assessment or other audit processes.

This document specifies the identification and the measurement of the activity in soils of a large number of gamma-emitting radionuclides using gamma spectrometry. This non-destructive method, applicable to large-volume samples (up to about 3 l), covers the determination in a single measurement of all the γ-emitters present for which the photon energy is between 5 keV and 3 MeV.
Generic test method and fundamentals using gamma-ray spectrometry are described in ISO 20042.
This document can be applied by test laboratories performing routine radioactivity measurements as a majority of gamma-emitting radionuclides is characterized by gamma-ray emission between 40 keV and 2 MeV.
The method can be implemented using a germanium or other type of detector with a resolution better than 5 keV.
This document addresses methods and practices for determining gamma-emitting radionuclides activity present in soil, including rock from bedrock and ore, construction materials and products, pottery, etc. This includes such soils and material containing naturally occurring radioactive material (NORM) or those from technological processes involving Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM) (e.g. the mining and processing of mineral sands or phosphate fertilizer production and use) as well as of sludge and sediment. This determination of gamma-emitting radionuclides activity is typically performed for the purpose of radiation protection. It is suitable for the surveillance of the environment and the inspection of a site and allows, in case of accidents, a quick evaluation of gamma activity of soil samples. This might concern soils from gardens, farmland, urban or industrial sites that can contain building materials rubble, as well as soil not affected by human activities.
When the radioactivity characterization of the unsieved material above 200 μm or 250 μm, made of petrographic nature or of anthropogenic origin such as building materials rubble, is required, this material can be crushed in order to obtain a homogeneous sample for testing as described in ISO 18589‑2.

This document provides criteria for quality assurance and quality control, evaluation of the performance and the accreditation of biological dosimetry by cytogenetic service laboratories using the dicentric assay performed with manual scoring.
This document is applicable to
a)    the confidentiality of personal information, for the requestor and the service laboratory,
b)    the laboratory safety requirements,
c)    the calibration sources and calibration dose ranges useful for establishing the reference dose-response curves that contribute to the dose estimation from unstable chromosome aberration frequency and the detection limit,
d)    the scoring procedure for unstable chromosome aberrations used for biological dosimetry,
e)    the criteria for converting a measured aberration frequency into an estimate of absorbed dose,
f)     the reporting of results,
g)    the quality assurance and quality control, and
h)    informative annexes containing sample instructions for requestor (see Annex A), sample questionnaire (see Annex B), sample report (see Annex C), fitting of the low dose-response curve by the method of maximum likelihood and calculating the error of the dose estimate (see Annex D), odds ratio method for cases of suspected exposure to a low dose (see Annex E), a method for determining the decision threshold and detection limit (see Annex F) and sample data sheet for recording aberrations (see Annex G).

This document specifies the neutron reference radiation fields, in the energy range from thermal up to 20 MeV, for calibrating neutron-measuring devices used for radiation protection purposes and for determining their response as a function of neutron energy.
This document is concerned only with the methods of producing and characterizing the neutron reference radiation fields. The procedures for applying these radiation fields for calibrations are described in References [1] and [2].
The neutron reference radiation fields specified are the following:
—    neutron fields from radionuclide sources, including neutron fields from sources in a moderator;
—    neutron fields produced by nuclear reactions with charged particles from accelerators;
—    neutron fields from reactors.
In view of the methods of production and use of them, these neutron reference radiation fields are divided, for the purposes of this document, into the following three separate clauses:
—    In Clause 4, radionuclide neutron sources with wide spectra are specified for the calibration of neutron-measuring devices. These sources should be used by laboratories engaged in the routine calibration of neutron-measuring devices, the particular design of which has already been type tested.
—    In Clause 5, accelerator-produced monoenergetic neutrons and reactor-produced neutrons with wide or quasi monoenergetic spectra are specified for determining the response of neutron‑measuring devices as a function of neutron energy. Since these neutron reference radiation fields are produced at specialized and well-equipped laboratories, only the minimum of experimental detail is given.
—    In Clause 6, thermal neutron fields are specified. These fields can be produced by moderated radionuclide sources, accelerators, or reactors.

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.

This document provides performance and test requirements for determining the acceptability of
neutron dosimetry systems to be used for the measurement of personal dose equivalent, Hp(10), for
neutrons ranging in energy from thermal to 20 MeV1).
This document applies to all passive neutron detectors that can be used within a personal dosemeter
in part or in all of the above-mentioned neutron energy range. No distinction between the different
techniques available in the marketplace is made in the description of the tests. Only generic distinctions,
for instance, as disposable or reusable dosemeters, are considered.
This document describes type tests only. Type tests are made to assess the basic characteristics of the
dosimetry systems and are often ensured by recognized national laboratories
This document does not present performance tests for characterizing the degradation induced by the
following:
— intrinsic temporal variability of the quality of the dosemeter supplied by the manufacturer;
— intrinsic temporal variability of preparation treatments (before irradiation and/or before reading),
if existing;
— intrinsic temporal variability of reading process;
— degradation due to environmental effects on the preparation treatments, if existing;
— degradation due to environmental effects on the reading process.

This document specifies a procedure, in the field of ionizing radiation metrology, for the calculation of the “decision threshold”, the “detection limit” and the “limits of the coverage interval” for a non negative ionizing radiation measurand when counting measurements with preselection of time or counts are carried out. The measurand results from a gross count rate and a background count rate as well as from further quantities on the basis of a model of the evaluation. In particular, the measurand can be the net count rate as the difference of the gross count rate and the background count rate, or the net activity of a sample. It can also be influenced by calibration of the measuring system, by sample treatment and by other factors.
ISO 11929-4 gives guidance to the application of ISO 11929 (all parts), summarizing shortly the general procedure and then presenting a wide range of numerical examples. The examples cover elementary applications according to ISO 11929-1 and ISO 11929-2.

This document provides methodology and criteria to qualify the dosimetry system at workplaces where
it is used. The criteria in this document apply to dosimetry systems which do not meet the criteria with
regard to energy and direction dependent responses described in ISO 21909-1.
The qualification of the dosimetry system at workplace aims to demonstrate that:
— either, the non-conformity of the dosimetry system to some of the requirements on the energy or
direction dependent responses defined in ISO 21909-1 does not lead to significant discrepancies in
the dose determination for a certain workplace field;
— or, that the correction factor or function used for this specific studied workplace enables the
dosimetry system to accurately determine the conventional dose value with uncertainties similar
to the ones given in ISO 21909-1.
The methodologies to characterize the work place field in order to perform the qualification of the
dosimetry system are given in Annex A. Annex B is complementary as it gives the practical methods to
follow, once one methodology is chosen.
The provider of the dosimetry system shall provide the type test results corresponding to ISO 21909-1.
However, when the dosimetry system to be qualified does not comply with all the criteria of ISO 21909-1
dealing with the energy and angle dependence of the response, some tests of the ISO 21909-1 can be not
performed.

This document specifies the characteristics of solid, liquid or gas sources of gamma emitting
radionuclides used as reference measurement standards for the calibration of gamma-ray spectrometers.
These reference measurement standards are traceable to national measurement standards.
This document does not describe the procedures involved in the use of these reference measurement
standards for the calibration of gamma-ray spectrometers. Such procedures are specified in ISO 20042
and other documents.
This document specifies recommended reference radiations for the calibration of gamma-ray
spectrometers. This document covers, but is not restricted to, gamma emitters which emit photons in
the energy range of 60 keV to 1 836 keV. These reference radiations are realized in the form of point
sources or adequately extended sources specified in terms of activity which are traceable to national
standards.

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

This document gives the basis for the measurement of ambient dose equivalent at flight altitudes for the evaluation of the exposures to cosmic radiation in civilian aircraft.

This document specifies minimum nuclear criticality safety training requirements for operations staff, operations supervisors, and management.
This document is applicable to areas, processes or facilities containing quantities of fissile material for which nuclear criticality safety assessment is required as defined in ISO 1709.
This document is not applicable to the transport of fissile materials outside the boundaries of nuclear establishments.

This document focuses on monitoring the activity concentrations of radioactive gases. They allow the calculation of the activity releases, in the gaseous effluent discharge from facilities producing positron emitting radionuclides and radiopharmaceuticals. Such facilities produce short-lived radionuclides used for medical purposes or research and can release gases typically including, but not limited to 18F, 11C, 15O and 13N. These facilities include accelerators, radiopharmacies, hospitals and universities. This document provides performance‑based criteria for the design and use of air monitoring equipment including probes, transport lines, sample monitoring instruments, and gas flow measuring methods. This document also provides information on monitoring program objectives, quality assurance, development of air monitoring control action levels, system optimisation and system performance verification.
The goal of achieving an unbiased measurement is accomplished either by direct (in-line) measurement on the exhaust stream or with samples extracted from the exhaust stream (bypass), provided that the radioactive gases are well mixed in the airstream. This document sets forth performance criteria and recommendations to assist in obtaining valid measurements.
NOTE 1 The criteria and recommendations of this document are aimed at monitoring which is conducted for regulatory compliance and system control. If existing air monitoring systems were not designed according to the performance criteria and recommendations of this document, an evaluation of the performance of the system is advised. If deficiencies are discovered based on a performance evaluation, a determination of the need for a system retrofit is to be made and corrective actions adopted where practicable.
NOTE 2 The criteria and recommendations of this document apply under both normal and off‑normal operating conditions, provided that these conditions do not include production of aerosols or vapours. If the normal and/or off-normal conditions produce aerosols and vapours, then the aerosol collection principles of ISO 2889 also apply.

The primary purpose of this document is to provide minimum acceptable criteria required to establish a procedure for retrospective dosimetry by electron paramagnetic resonance spectroscopy and to report the results.
The second purpose is to facilitate the comparison of measurements related to absorbed dose estimation obtained in different laboratories.
This document covers the determination of absorbed dose in the measured material. It does not cover the calculation of dose to organs or to the body. It covers measurements in both biological and inanimate samples, and specifically:
a)   based on inanimate environmental materials like glass, plastics, clothing fabrics, saccharides, etc., usually made at X-band microwave frequencies (8 GHz to 12 GHz);
b)   in vitro tooth enamel using concentrated enamel in a sample tube, usually employing X-band frequency, but higher frequencies are also being considered;
c)   in vivo tooth dosimetry, currently using L-band (1 GHz to 2 GHz), but higher frequencies are also being considered;
d)   in vitro nail dosimetry using nail clippings measured principally at X-band, but higher frequencies are also being considered;
e)   in vivo nail dosimetry with the measurements made at X-band on the intact finger or toe;
f)    in vitro measurements of bone, usually employing X-band frequency, but higher frequencies are also being considered.
For biological samples, in vitro measurements are carried out in samples after their removal from the person or animal and under laboratory conditions, whereas the measurements in vivo are carried out without sample removal and may take place under field conditions.
NOTE    The dose referred to in this document is the absorbed dose of ionizing radiation in the measured materials.

This document specifies the different leakage test methods for sealed sources. It gives a comprehensive set of procedures using radioactive and non-radioactive means.
This document applies to the following situations:
—     leakage testing of test sources following design classification testing in accordance with ISO 2919[1];
—     production quality control testing of sealed sources;
—     periodic inspections of the sealed sources performed at regular intervals, during the working life.
Annex A of this document gives guidance to the user in the choice of the most suitable method(s) according to situation and source type.
It is recognized that there can be circumstances where special tests, not described in this document, are required.
It is emphasized, however, that insofar as production, use, storage and transport of sealed radioactive sources are concerned, compliance with this document is no substitute for complying with the requirements of the relevant IAEA regulations[17] and other relevant national regulations. It is also recognized that countries can enact statutory regulations which specify exemptions for tests, according to sealed source type, design, working environment, and activity (e.g., for very low activity reference sources where the total activity is less than the leakage test limit).

This document specifies the characteristics of reference measurement standards of radioactive surface contamination, traceable to national measurement standards, for the calibration of surface contamination monitors. This document relates to alpha-emitters, beta-emitters, and photon emitters of maximum photon energy not greater than 1,5 MeV.
It does not describe the procedures involved in the use of these reference measurement standards for the calibration of surface contamination monitors. Such procedures are specified in IEC 60325[6], IEC 62363[7], and other documents.
NOTE    Since some of the proposed photon standards include filters, the photon standards are to be regarded as reference measurement standards of photons of a particular energy range and not as reference measurement standards of a particular radionuclide. For example, a 241Am source with the recommended filtration does not emit from the surface the alpha particles or characteristic low-energy L X-ray photons associated with the decay of the nuclide. It is designed to be a reference measurement standard that emits photons with an average energy of approximately 60 keV.
This document also specifies preferred reference radiations for the calibration of surface contamination monitors. These reference radiations are realized in the form of adequately characterized large area sources specified, without exception, in terms of surface emission rate and activity which are traceable to national standards.

The purpose of this document is to provide minimum criteria required for quality assurance and quality control, evaluation of the performance and to facilitate the comparison of measurements related to absorbed dose estimation obtained in different laboratories applying ex vivo X-band EPR spectroscopy with human tooth enamel.
This document covers the determination of absorbed dose in tooth enamel (hydroxyapatite). It does not cover the calculation of dose to organs or to the body.
This document addresses:
a)   responsibilities of the customer and laboratory;
b)   confidentiality and ethical considerations;
c)   laboratory safety requirements;
d)   the measurement apparatus;
e)   preparation of samples;
f)    measurement of samples and EPR signal evaluation;
g)   calibration of EPR dose response;
h)   dose uncertainty and performance test;
i)     quality assurance and control.

This document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements.
It does not add to, subtract from, or in any way modify those requirements.
This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.

This document specifies the minimum requirements for the design of professional programmes to monitor workers exposed to a risk of ingestion to uranium compounds. This document establishes principles for the development of compatible goals and requirements for monitoring programmes and dose assessment for workers occupationally exposed to internal contamination. It establishes procedures and assumptions for risk analysis, monitoring programmes and the standardized interpretation of monitoring data in order to achieve acceptable levels of reliability for uranium and its compounds. It sets limits for the applicability of the procedures in respect to dose levels above which more sophisticated methods need to be applied.
This document addresses those circumstances when exposure could be constrained by either radiological or chemical toxicity concerns.
This document addresses, for ingestion of uranium and its compounds, the following items:
a)   purposes of monitoring and monitoring programmes;
b)   description of the different categories of monitoring programmes;
c)   suitable methods for monitoring and criteria for their selection;
d)   information that is collected for the design of a monitoring programme;
e)   procedures for dose assessment based on reference levels for special monitoring programmes;
f)    criteria for determining the significance of monitoring results;
g)   uncertainties arising from dose assessment and interpretation of bioassays data;
h)   reporting/documentation;
i)    quality assurance;
j)    record keeping requirements.
It is not applicable to the following items:
a)   detailed descriptions of measuring methods and techniques for uranium;
b)   modelling for the improvement of internal dosimetry;
c)   potential influence of counter-measures (e.g. administration of chelating agents);
d)   investigation of the causes or implications of an exposure;
e)   dosimetry for inhalation exposures and for contaminated wounds.

This International Standard specifies requirements for a quality management system when an organization:
a)    needs to demonstrate its ability to consistently provide products and services that meet customer and applicable statutory and regulatory requirements, and
b)    aims to enhance customer satisfaction through the effective application of the system, including processes for improvement of the system and the assurance of conformity to customer and applicable statutory and regulatory requirements.
All the requirements of this International Standard are generic and are intended to be applicable to any organization, regardless of its type or size, or the products and services it provides.
NOTE 1    In this International Standard, the terms "product" or "service" only apply to products and services intended for, or required by, a customer.
NOTE 2    Statutory and regulatory requirements can be expressed as legal requirements.
This International Standard applies to organizations supplying ITNS products or services.
Application of this standard to organizations performing activities on a licensed nuclear site is subject to prior agreement by the Licensee.
Requirements specified in this International Standard are complementary (not alternative) to customer and applicable statutory and regulatory requirements.

This document specifies the requirements for personal contamination monitoring and dose assessment following wounds involving radioactive materials. It includes requirements for the direct monitoring at the wound site, monitoring of uptake of radionuclides into the body and assessment of local and systemic doses following the wound event.
It does not address:
—     details of monitoring and assessment methods for specific radionuclides;
—     monitoring and dose assessment for materials in contact with intact skin or pre-existing wounds, including hot particles;
—     therapeutic protocols. However, the responsible entity needs to address the requirements for decontamination and decorporation treatments if appropriate.

This document provides a method that allows an estimation of gross radioactivity of alpha- and beta-emitters present in soil samples. It applies, essentially, to systematic inspections based on comparative measurements or to preliminary site studies to guide the testing staff both in the choice of soil samples for measurement as a priority and in the specific analysis methods for implementation.
The gross α or β radioactivity is generally different from the sum of the effective radioactivities of the radionuclides present since, by convention, the same alpha counting efficiency is assigned for all the alpha emissions and the same beta counting efficiency is assigned for all the beta emissions.
Soil includes rock from bedrock and ore as well as construction materials and products, potery, etc. using naturally occurring radioactive materials (NORM) or those from technological processes involving Technologically Enhanced Naturally Occurring Radioactive Materials (TENORM), e.g. the mining and processing of mineral sands or phosphate fertilizer production and use.
The test methods described in this document can also be used to assess gross radioactivity of alpha- and beta-emitters in sludge, sediment, construction material and products following proper sampling procedure[2][3][4][5][7][8].
For simplification, the term "soil" used in this document covers the set of elements mentioned above.

This document specifies the requirements applicable to the design and use of airborne confinement systems that ensure safety and radioprotection functions in nuclear worksites and in nuclear installations under decommissioning to protect from radioactive contamination produced: aerosol or gas.
The purpose of confinement systems is to protect the workers, members of the public and environment against the spread of radioactive contamination resulting from operations in nuclear worksites and from nuclear installations under decommissioning.
The confinement of nuclear worksites and of nuclear installations under decommissioning is characterized by the temporary and evolving (dynamic) nature of the operations to be performed. These operations often take place in area not specifically designed for this purpose.
This document applies to maintenance or upgrades at worksites which fit the above definition.
NOTE       The requirements for the design and use of ventilation and confinement systems and for liquid confinement in nuclear reactors or in nuclear installations other than nuclear worksites and nuclear installations under decommissioning are developed in other ISO standards.

This document describes a method for measuring 238Pu and 239 + 240 isotopes in soil by alpha spectrometry samples using chemical separation techniques.
The method can be used for any type of environmental study or monitoring. These techniques can also be used for measurements of very low levels of activity, one or two orders of magnitude less than the level of natural alpha-emitting radionuclides.
The test methods described in this document can also be used to measure the radionuclides in sludge, sediment, construction material and products following proper sampling procedure[2][3][4][5][7][8].
The mass of the test portion required depends on the assumed activity of the sample and the desired detection limit. In practice, it can range from 0,1 g to 100 g of the test sample.

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.

This document describes the principles for the measurement of the activity of 90Sr in equilibrium with 90Y and 89Sr, pure beta emitting radionuclides, in soil samples. Different chemical separation methods are presented to produce strontium and yttrium sources, the activity of which is determined using proportional counters (PC) or liquid scintillation counters (LSC). 90Sr can be obtained from the test samples when the equilibrium between 90Sr and 90Y is reached or through direct 90Y measurement. The selection of the measuring method depends on the origin of the contamination, the characteristics of the soil to be analysed, the required accuracy of measurement and the resources of the available laboratories.
These methods are used for soil monitoring following discharges, whether past or present, accidental or routine, liquid or gaseous. It also covers the monitoring of contamination caused by global nuclear fallout.
In case of recent fallout immediately following a nuclear accident, the contribution of 89Sr to the total amount of strontium activity will not be negligible. This standard provides the measurement method to determine the activity of 90Sr in presence of 89Sr.
The test methods described in this document can also be used to measure the radionuclides in sludge, sediment, construction material and products by following proper sampling procedure.
Using samples sizes of 20 g and counting times of 1 000 min, detection limits of (0,1 to 0,5) Bq·kg-1 can be achievable for 90Sr using conventional and commercially available proportional counter or liquid scintillation counter when the presence of 89Sr can be neglected. If 89Sr is present in the test sample, detection limits of (1 to 2) Bq·kg-1 can be obtained for both 90Sr and 89Sr using the same sample size, counting time and proportional counter or liquid scintillation counter as in the previous situation.