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

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 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 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 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 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 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 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 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 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 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 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 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 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.

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 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 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.

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 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.

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 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 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 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.

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 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 describes the methods for determining the activity in becquerel (Bq) of gamma‑ray emitting radionuclides in test samples by gamma-ray spectrometry. The measurements are carried out in a testing laboratory following proper sample preparation. The test samples can be solid, liquid or gaseous. Applications include:
—          routine surveillance of radioactivity released from nuclear installations or from sites discharging enhanced levels of naturally occurring radioactive materials;
—          contributing to determining the evolution of radioactivity in the environment;
—          investigating accident and incident situations, in order to plan remedial actions and monitor their effectiveness;
—          assessment of potentially contaminated waste materials from nuclear decommissioning activities;
—          surveillance of radioactive contamination in media such as soils, foodstuffs, potable water, groundwaters, seawater or sewage sludge;
—          measurements for estimating the intake (inhalation, ingestion or injection) of activity of gamma-ray emitting radionuclides in the body.
It is assumed that the user of this document has been given information on the composition of the test sample or the site. In some cases, the radionuclides for analysis have also been specified if characteristic limits are needed. It is also assumed that the test sample has been homogenised and is representative of the material under test.
General guidance is included for preparing the samples for measurement. However, some types of sample are to be prepared following the requirements of specific standards referred to in this document. The generic recommendations can also be useful for the measurement of gamma-ray emitters in situ.
This document includes generic advice on equipment selection (see Annex A), detectors (more detailed information is included in Annex D), and commissioning of instrumentation and method validation. Annex F summarises the influence of different measurement parameters on results for a typical gamma-ray spectrometry system. Quality control and routine maintenance are also covered, but electrical testing of the detector and pulse processing electronics is excluded. It is assumed that any data collection and analysis software used has been written and tested in accordance with relevant software standards such as ISO/IEC/IEEE 12207.
Calibration using reference sources and/or numerical methods is covered, including verification of the results. It also covers the procedure to estimate the activity content of the sample (Bq) from the spectrum.
The principles set out in this document are applicable to measurements by gamma-ray spectrometry in testing laboratories and in situ. However, the detailed requirements for in situ measurement are given in ISO 18589-7 and are outside the scope of this document.
This document covers, but is not restricted to, gamma-ray emitters which emit photons in the energy range of 5 keV to 3 000 keV. However, most of the measurements fall into the range 40 keV to 2 000 keV. The activity (Bq) ranges from the low levels (sub-Bq) found in environmental samples to activities found in accident conditions and high level radioactive wastes.

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 specifies the general requirements to carry out radionuclides tests, including sampling of soil including rock from bedrock and ore as well as of construction materials and products, pottery, etc. using 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.
For simplification, the term "soil" used in this document covers the set of elements mentioned above.
This document is addressed to people responsible for determining the radioactivity present in soils for the purpose of radiation protection. This concerns soils from gardens and farmland, urban or industrial sites, as well as soil not affected by human activities.
This document is applicable to all laboratories regardless of the number of personnel or the extent of the scope of testing activities. When a laboratory does not undertake one or more of the activities covered by this document, such as planning, sampling or testing, the requirements of those clauses do not apply.
This document is to be used in conjunction with other parts of ISO 18589 that outline the setting up of programmes and sampling techniques, methods of general processing of samples in the laboratory and also methods for measuring the radioactivity in soil. Its purpose is the following:
—     define the main terms relating to soils, sampling, radioactivity and its measurement;
—     describe the origins of the radioactivity in soils;
—     define the main objectives of the study of radioactivity in soil samples;
—     present the principles of studies of soil radioactivity;
—     identify the analytical and procedural requirements when measuring radioactivity in soil.
This document is applicable if radionuclide measurements for the purpose of radiation protection are to be made in the following cases:
—     initial characterization of radioactivity in the environment;
—     routine surveillance of the impact of nuclear installations or of the evolution of the general territory;
—     investigations of accident and incident situations;
—     planning and surveillance of remedial action;
—     decommissioning of installations or clearance of materials.

This document is intended for the validation of codes used for the calculation of doses received by individuals on board aircraft. It gives guidance to radiation protection authorities and code developers on the basic functional requirements which the code fulfils.
Depending on any formal approval by a radiation protection authority, additional requirements concerning the software testing can apply.

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.

ISO 18229:2018 defines the essential technical requirements that are addressed in the process of design and construction of Generation IV (GEN IV) nuclear reactors. It does not address operation, maintenance and in-service inspection of reactors.
Six reactor concepts are considered for GEN IV: the sodium fast reactor, the lead fast reactor, the gas fast reactor, the very high temperature reactor, the supercritical water reactor and the molten salt reactor.
Annex A details the main characteristics for the different concepts.
The scope of application of this document is limited to mechanical components related to nuclear safety and to the prevention of the release of radioactive materials
?   that are considered to be important in terms of nuclear safety and operability,
?   that play a role in ensuring leaktightness, partitioning, guiding, securing and supporting, and
?   that contain and/or are in contact with fluids (such as vessels, pumps, valves, pipes, bellows, box structures, heat exchangers, handling and driving mechanisms).

This document specifies additional procedures and data for the calibration of dosemeters and doserate meters used for individual and area monitoring in radiation protection. The general procedure for the calibration and the determination of the response of radiation protection dose(rate)meters is described in ISO 29661 and is followed as far as possible. For this purpose, the photon reference radiation fields with mean energies between 8 keV and 9 MeV, as specified in ISO 4037-1, are used. In Annex D some additional information on reference conditions, required standard test conditions and effects associated with electron ranges are given. For individual monitoring, both whole body and extremity dosemeters are covered and for area monitoring, both portable and installed dose(rate)meters are covered.
Charged particle equilibrium is needed for the reference fields although this is not always established in the workplace fields for which the dosemeter should be calibrated. This is especially true at photon energies without inherent charged particle equilibrium at the reference depth d, which depends on the actual combination of energy and reference depth d. Electrons of energies above 65 keV, 0,75 MeV and 2,1 MeV can just penetrate 0,07 mm, 3 mm and 10 mm of ICRU tissue, respectively, and the radiation qualities with photon energies above these values are considered as radiation qualities without inherent charged particle equilibrium for the quantities defined at these depths. This document also deals with the determination of the response as a function of photon energy and angle of radiation incidence. Such measurements can represent part of a type test in the course of which the effect of further influence quantities on the response is examined.
This document is only applicable for air kerma rates above 1 µGy/h.
This document does not cover the in-situ calibration of fixed installed area dosemeters.
The procedures to be followed for the different types of dosemeters are described. Recommendations are given on the phantom to be used and on the conversion coefficients to be applied. Recommended conversion coefficients are only given for matched reference radiation fields, which are specified in ISO 4037-1:2019, Clauses 4 to 6. ISO 4037‑1:2019, Annexes A and B, both informative, include fluorescent radiations, the gamma radiation of the radionuclide 241Am, S-Am, for which detailed published information is not available. ISO 4037-1:2019, Annex C, gives additional X radiation fields, which are specified by the quality index. For all these radiation qualities, conversion coefficients are given in Annexes A to C, but only as a rough estimate as the overall uncertainty of these conversion coefficients in practical reference radiation fields is not known.
NOTE       The term dosemeter is used as a generic term denoting any dose or doserate meter for individual or area monitoring.

This document describes rules for the procedures, applications, and systems of thermoluminescence dosimetry (TLD) for dose measurements according to the probe method. It is particularly applicable to solid "TL detectors", i.e. rods, chips, and microcubes, made from LiF:Mg,Ti or LiF:Mg,Cu,P in crystalline or polycrystalline form. It is not applicable to LiF powders because their use requires special procedures. The probe method encompasses the arrangement, particularly in a water phantom or in a tissue-equivalent phantom, of single TL detectors or of "TL probes", i.e. sets of TL detectors arranged in thin-walled polymethyl methacrylate (PMMA) casings.
The purpose of these rules is to guarantee the reliability and the accuracy indispensable in clinical dosimetry when applied on or in the patient or phantom. This document applies to dosimetry in teletherapy with both photon radiation from 20 keV to 50 MeV and electron radiation from 4 MeV to 25 MeV, as well as in brachytherapy with photon-emitting radionuclides. These applications are complementary to the use of ionization chambers.

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.

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

This document specifies the procedures for the dosimetry of X and gamma reference radiation for the calibration of radiation protection instruments over the energy range from approximately 8 keV to 1,3 MeV and from 4 MeV to 9 MeV and for air kerma rates above 1 µGy/h. The considered measuring quantities are the air kerma free-in-air, Ka, and the phantom related operational quantities of the International Commission on Radiation Units and Measurements (ICRU)[2], H*(10), Hp(10), H'(3), Hp(3), H'(0,07) and Hp(0,07), together with the respective dose rates. The methods of production are given in ISO 4037-1.
This document can also be used for the radiation qualities specified in ISO 4037-1:2019, Annexes A, B and C, but this does not mean that a calibration certificate for radiation qualities described in these annexes is in conformity with the requirements of ISO 4037.
The requirements and methods given in this document are targeted at an overall uncertainty (k = 2) of the dose(rate) of about 6 % to 10 % for the phantom related operational quantities in the reference fields. To achieve this, two production methods of the reference fields are proposed in ISO 4037-1.
The first is to produce "matched reference fields", which follow the requirements so closely that recommended conversion coefficients can be used. The existence of only a small difference in the spectral distribution of the "matched reference field" compared to the nominal reference field is validated by procedures, which are given and described in detail in this document. For matched reference radiation fields, recommended conversion coefficients are given in ISO 4037-3 only for specified distances between source and dosemeter, e.g., 1,0 m and 2,5 m. For other distances, the user has to decide if these conversion coefficients can be used.
The second method is to produce "characterized reference fields". Either this is done by determining the conversion coefficients using spectrometry, or the required value is measured directly using secondary standard dosimeters. This method applies to any radiation quality, for any measuring quantity and, if applicable, for any phantom and angle of radiation incidence. The conversion coefficients can be determined for any distance, provided the air kerma rate is not below 1 µGy/h.
Both methods require charged particle equilibrium for the reference field. However this is not always established in the workplace field for which the dosemeter shall be calibrated. This is especially true at photon energies without inherent charged particle equilibrium at the reference depth d, which depends on the actual combination of energy and reference depth d. Electrons of energies above 65 keV, 0,75 MeV and 2,1 MeV can just penetrate 0,07 mm, 3 mm and 10 mm of ICRU tissue, respectively, and the radiation qualities with photon energies above these values are considered as radiation qualities without inherent charged particle equilibrium for the quantities defined at these depths.
This document is not applicable for the dosimetry of pulsed reference fields.

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.

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.

This document specifies the characteristics and production methods of X and gamma reference radiation for calibrating protection-level dosemeters and doserate meters with respect to the phantom related operational quantities of the International Commission on Radiation Units and Measurements (ICRU)[5]. The lowest air kerma rate for which this standard is applicable is 1 µGy h?1. Below this air kerma rate the (natural) background radiation needs special consideration and this is not included in this document.
For the radiation qualities specified in Clauses 4 to 6, sufficient published information is available to specify the requirements for all relevant parameters of the matched or characterized reference fields in order to achieve the targeted overall uncertainty (k = 2) of about 6 % to 10 % for the phantom related operational quantities. The X ray radiation fields described in the informative Annexes A to C are not designated as reference X-radiation fields.
NOTE       The first edition of ISO 4037-1, issued in 1996, included some additional radiation qualities for which such published information is not available. These are fluorescent radiations, the gamma radiation of the radionuclide 241Am, S-Am, and the high energy photon radiations R-Ti and R-Ni, which have been removed from the main part of this document. The most widely used radiations, the fluorescent radiations and the gamma radiation of the radionuclide 241Am, S-Am, are included nearly unchanged in the informative Annexes A and B. The informative Annex C gives additional X radiation fields, which are specified by the quality index.
The methods for producing a group of reference radiations for a particular photon-energy range are described in Clauses 4 to 6, which define the characteristics of these radiations. The three groups of reference radiation are:
a)    in the energy range from about 8 keV to 330 keV, continuous filtered X radiation;
b)    in the energy range 600 keV to 1,3 MeV, gamma radiation emitted by radionuclides;
c)    in the energy range 4 MeV to 9 MeV, photon radiation produced by accelerators.
The reference radiation field most suitable for the intended application can be selected from Table 1, which gives an overview of all reference radiation qualities specified in Clauses 4 to 6. It does not include the radiations specified in the Annexes A, B and C.
The requirements and methods given in Clauses 4 to 6 are targeted at an overall uncertainty (k = 2) of the dose(rate) value of about 6 % to 10 % for the phantom related operational quantities in the reference fields. To achieve this, two production methods are proposed:
The first one is to produce "matched reference fields", whose properties are sufficiently well-characterized so as to allow the use of the conversion coefficients recommended in ISO 4037-3. The existence of only a small difference in the spectral distribution of the "matched reference field" compared to the nominal reference field is validated by procedures, which are given and described in detail in ISO 4037‑2. For matched reference radiation fields, recommended conversion coefficients are given in ISO 4037‑3 only for specified distances between source and dosemeter, e.g., 1,0 m and 2,5 m. For other distances, the user has to decide if these conversion coefficients can be used. If both values are very similar, e.g., differ only by 2 % or less, then a linear interpolation may be used.
The second method is to produce "characterized reference f

The ISO 11929 series 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 has been divided into four parts covering elementary applications in ISO 11929-1, advanced applications on the basis of the GUM Supplement 1 in this document, applications to unfolding methods in ISO 11929-3, and guidance to the application in ISO 11929-4.
ISO 11929-1 covers basic applications of counting measurements frequently used in the field of ionizing radiation metrology. It is restricted to applications for which the uncertainties can be evaluated on the basis of the ISO/IEC Guide 98-3 (JCGM 2008). In Annex A of ISO 11929-1:2019 the special case of repeated counting measurements with random influences is covered, while measurements with linear analogous ratemeters are covered in Annex B of ISO 11929-1:2019.
This document extends the former ISO 11929:2010 to the evaluation of measurement uncertainties according to the ISO/IEC Guide 98-3-1. It also presents some explanatory notes regarding general aspects of counting measurements and on Bayesian statistics in measurements.
ISO 11929-3 deals with the evaluation of measurements using unfolding methods and counting spectrometric multi-channel measurements if evaluated by unfolding methods, in particular, for alpha- and gamma‑spectrometric measurements. Further, it provides some advice on how to deal with correlations and covariances.
ISO 11929-4 gives guidance to the application of ISO 11929, summarizes shortly the general procedure and then presents a wide range of numerical examples. Information on the statistical roots of ISO 11929 and on its current development may be found elsewhere[30,31].
ISO 11929 also applies analogously to other measurements of any kind especially if a similar model of the evaluation is involved. Further practical examples can be found, for example, in ISO 18589[1], ISO 9696[2], ISO 9697[3], ISO 9698[4], ISO 10703[5], ISO 7503[6], ISO 28218[7], and ISO 11885[8].
NOTE       A code system, named UncertRadio, is available for calculations according to ISO 119291 to ISO 11929-3. UncertRadio[27][28] can be downloaded for free from https://www.thuenen.de/en/fi/fields-of-activity/marine-environment/coordination-centre-of-radioactivity/uncertradio/. The download contains a setup installation file which copies all files and folders into a folder specified by the user. After installation one has to add information to the PATH of Windows as indicated by a pop‑up window during installation. English language can be chosen and extensive "help" information is available. . Another tool is the package ?metRology'[32] which is available for programming in R. It contains the two R functions ?uncert' and ?uncertMC' which perform the GUM conform uncertainty propagation, either analytically or by the Monte Carlo method, respectivel

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