ISO/TC 85/SC 2 - Radiological protection

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 ISO 11929-1:2025, Annex A, the special case of repeated counting measurements with random influences is covered, while measurements with linear analogous ratemeters are covered in ISO 11929-1:2025, Annex B. 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[ REF Reference_ref_37 \r \h 30 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00330037000000 ][ REF Reference_ref_38 \r \h 31 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00330038000000 ]. 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[ REF Reference_ref_8 \r \h 1 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0038000000 ], ISO 9696[ REF Reference_ref_9 \r \h 2 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0039000000 ], ISO 9697[ REF Reference_ref_10 \r \h 3 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310030000000 ], ISO 9698[ REF Reference_ref_11 \r \h 4 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310031000000 ], ISO 10703[ REF Reference_ref_12 \r \h 5 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310032000000 ], ISO 7503[ REF Reference_ref_13 \r \h 6 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310033000000 ], ISO 28218[ REF Reference_ref_14 \r \h 7 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310034000000 ] and ISO 11665[ REF Reference_ref_15 \r \h 8 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310035000000 ]. NOTE A code system, named UncertRadio, is available for calculations according t

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 ISO/IEC Guide 98-3:2008/Suppl 1:2008 in ISO 11929-2, applications to unfolding methods in this document, 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 ISO 11929-1:2025, Annex A, the special case of repeated counting measurements with random influences is covered, while measurements with linear analogous ratemeters, are covered in ISO 11929-1:2025, Annex B. This document 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 the ISO 11929 series, summarizes shortly the general procedure and then presents a wide range of numerical examples. ISO 11929 Standard 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[ REF Reference_ref_14 \r \h 7 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310034000000 ], ISO 9696[ REF Reference_ref_9 \r \h 2 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0039000000 ], ISO 9697[ REF Reference_ref_10 \r \h 3 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310030000000 ], ISO 9698[ REF Reference_ref_11 \r \h 4 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310031000000 ], ISO 10703[ REF Reference_ref_12 \r \h 5 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310032000000 ], ISO 7503[ REF Reference_ref_8 \r \h 1 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0038000000 ], ISO 28218[ REF Reference_ref_15 \r \h 8 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310035000000 ], and ISO 11665[ REF Reference_ref_13 \r \h 6 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310033000000 ]. NOTE A code system, named UncertRadio, is available for calculations according to ISO 11929- 1 to ISO 11929-3. UncertRadio[ REF Reference_ref_42 \r \h 35 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00340032000000 ][ REF Reference_ref_43 \r \h 36 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00340033000000 ] can be downloaded

This document specifies the identification of radionuclides and the measurement of their activity in soil using in situ gamma spectrometry with portable systems equipped with germanium or scintillation detectors. This document is suitable to rapidly assess the activity of artificial and natural radionuclides deposited on or present in soil layers of large areas of a site under investigation. This document can be used in connection with radionuclide measurements of soil samples in the laboratory (see ISO 18589-3) in the following cases: — routine surveillance of the impact of radioactivity released from nuclear installations or of the evolution of radioactivity in the region; — investigations of accident and incident situations; — planning and surveillance of remedial action; — decommissioning of installations or the clearance of materials. It can also be used for the identification of airborne artificial radionuclides, when assessing the exposure levels inside buildings or during waste disposal operations. Following a nuclear accident, in situ gamma spectrometry is a powerful method for rapid evaluation of the gamma activity deposited onto the soil surface as well as the surficial contamination of flat objects. NOTE The method described in this document is not suitable when the spatial distribution of the radionuclides in the environment is not precisely known (influence quantities, unknown distribution in soil) or in situations with very high photon flux. However, the use of small volume detectors with suitable electronics allows measurements to be performed under high photon flux.

This document specifies approaches for the determination of plutonium isotopes (238Pu, 239Pu and 240Pu) in urine using alpha spectrometry or inductively coupled plasma mass spectrometry (ICP-MS). It is applicable to the measurement of plutonium isotopes at levels which are appropriate for — workers handling plutonium in planned exposure situations, where detection limits are sufficiently low to be in accordance with dose limits, and — workers, members of the public and emergency responders in emergency exposure situations, where required detection limits can be much higher, and results need to be reported in a short timescale. This document does not provide information on when monitoring is carried out or the interpretation of the results in terms of dose or biological effects.

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 this document, advanced applications on the basis of the ISO/IEC Guide 3-1 in ISO 11929-2, applications to unfolding methods in ISO 11929-3, and guidance to the application in ISO 11929-4. This document 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 REF Annex_sec_A \r \h Annex A 08D0C9EA79F9BACE118C8200AA004BA90B02000000080000000C00000041006E006E00650078005F007300650063005F0041000000 , the special case of repeated counting measurements with random influences is covered, while measurements with linear analogous ratemeters are covered in REF Annex_sec_B \r \h Annex B 08D0C9EA79F9BACE118C8200AA004BA90B02000000080000000C00000041006E006E00650078005F007300650063005F0042000000 . ISO 11929-2 extends the former ISO 11929:2010 to the evaluation of measurement uncertainties according to the ISO/IEC Guide 98-3:2008/Suppl 1:2008. ISO 11929-2 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 the ISO 11929 series, 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[ REF Reference_ref_40 \r \h 33 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00340030000000 ][ REF Reference_ref_41 \r \h 34 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00340031000000 ]. The ISO 11929 series 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[ REF Reference_ref_8 \r \h 1 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0038000000 ], ISO 9696[ REF Reference_ref_9 \r \h 2 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0039000000 ], ISO 9697[ REF Reference_ref_10 \r \h 3 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310030000000 ], ISO 9698[ REF Reference_ref_11 \r \h 4 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310031000000 ], ISO 10703[ REF Reference_ref_12 \r \h 5 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310032000000 ], ISO 7503[ REF Reference_ref_13 \r \h 6 08D0C9EA79

This document specifies the different methods intended for assessing the radon diffusion coefficient in waterproofing materials such as bitumen or polymeric membranes, coatings or paints, as well as assumptions and boundary conditions that shall be met during the test. This document is not applicable for porous materials, where radon diffusion depends on porosity and moisture content.

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 [ REF Reference_ref_12 \r \h 9 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310032000000 ] [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.

This document specifies requirements concerning safety systems raised by the application of superconducting magnets in fusion facilities. Safety systems include confinement systems (both static and dynamic types), shielding barriers, penetrations, and supporting systems such as instrumentation and control. The requirements are applicable to both normal and abnormal operation of a fusion facility. For instance, the radiation protection shall be adequate in order to permit the hands-on operation to the electronics and parts for inspection, maintenance and replacement; the hazards associated with superconducting magnets, such as the loss of superconductivity (quench), Paschen breakdown following helium and voltage leakage, shall be prevented from breaching the integrity of safety systems. This document will facilitate the design and assessment of the safety systems in a fusion facility with superconducting magnets for all configurations, such as tokamak, stellarator and magneto-inertial fusion devices. Based on the advancement and maturity of the tokamak configuration, this document outlines safety requirements mostly derived from the tokamak configuration but also applicable to other configurations and layouts that may be adopted by future fusion devices.

This document is applicable to the radiation shielding design and evaluation work for medical proton accelerators of proton energies ranging from 70 MeV to 250 MeV, with subsystems such as beam transport system and nozzle components. The radiation protection recommendations given in this document cover the aspects relating to regulations, shielding design goals and other design criteria, role of the manufacturers, of the radiation protection officer or qualified expert, the medical physicist, the licensee and interactions between them, sources and radiations around a proton accelerator, shielding for accelerators and its subsystems (including shielding materials and transmission values, calculations for various room configurations, duct impact on radiation protection) and the radiological measurements. FLASH proton therapy is not covered by this document. NOTE 1 Annex A provides a list of the most used Monte-Carlo codes for shielding calculation. NOTE 2 Annex B provides the analytical methods and the corresponding necessary data for shielding calculation. NOTE 3 Annex C provides a set of examples on shielding calculation of barriers, maze and skyshine problems. NOTE 4 Annex D provides radiation shielding consideration on special topics.

The objective of this document is to characterize the gaseous effluents tritium and carbon-14 generated by nuclear facilities during operation and decommissioning and occurring in the same chemical species as hydrogen and carbon, e. g. as water vapour (HTO), hydrogen gas (HT, TT), carbon dioxide (14CO2), carbon monoxide (14CO), methane (CH3T, 14CH4). It concerns measurements on samples that are representative of a certain volume stream or volume of discharge during a given period of time and of the corresponding volume discharged. The result is therefore expressed in becquerels. This document applies to samples that were obtained by sampling methods according to ISO 20041-1[ REF Reference_ref_10 \r \h 9 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310030000000 ] and describes — analysis methods for the determination of tritium and carbon-14 activities by liquid scintillation counting, and — calculation methods to determine the tritium activities discharged as tritiated water vapour (HTO) and tritium in other chemical compounds (non-HTO) as well as carbon-14 activities discharged as carbon dioxide (14CO2) and carbon-14 in other chemical compounds (non-14CO2). This document does not apply to tritium and carbon-14 activity concentrations in the environmental air, e.g. in the vicinity of nuclear installations. The accountability rules of the activities discharged necessary for the establishment of regulatory reports do not fall within the scope of this document and are the responsibility of the regulatory bodies.

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 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 REF Section_sec_8.5 \r \h 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[ REF Reference_ref_8 \r \h 2 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000100000005200650066006500720065006E00630065005F007200650066005F0038000000 ]. 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].

The purpose of this document is to set out the general principles pertaining to the sampling strategy, to the collection and conditioning of samples, to their transport to the laboratory and to the pre-treatment operations to be carried out prior to analysis. It is intended for the use of organisations that implement a sampling programme as well as organisations responsible for collecting samples of bioindicators. These principles are not directly applicable to accident or post-accident situations. These principles can apply to biological matrices in the environment.

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 applies to the testing of surfaces that may become contaminated by radioactive materials. The ease of decontamination is a property of a surface and an important criterion for selecting surface materials used in the nuclear industry, interim storage or disposal facilities from which contamination can be removed easily and rapidly without damaging the surface. The test described in this document is a rapid laboratory-based method to compare the ease of decontamination of different surface materials. The results from the test can be one parameter to take into account when selecting surface coatings such as varnish or impervious layers such as ceramics and other surfaces. The radionuclides used in this test are those commonly found in the nuclear industry (137Cs, 134Cs and 60Co) in aqueous form. The test can also be adopted for use with other radionuclides and other chemical forms, depending on the customer requirements, if the solutions are chemically stable and do not corrode the test specimen. The test does not measure the ease of decontamination of the surface materials in practical use, as this depends on the radionuclide(s) present, their chemical form, the duration of exposure to the contaminant and the environmental conditions amongst other factors. The test method is not intended to describe general decontamination procedures or to assess the efficiency of decontamination procedures (see ISO 7503-1 to ISO 7503-3). The test method is not suitable for use of radiochemicals if the radionuclide emits low energy gamma rays or beta particles that are readily attenuated in the surface.

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 (References [4] and [5]) 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 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 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. The types of confinement systems for other facilities are covered by ISO 26802 for fission nuclear reactors, by ISO 17873 for facilities other than fission nuclear reactors and by ISO 16647 for nuclear worksite and for nuclear installations under decommissioning. The facilities covered by these three standards, notably ISO 17873, include tritium as a radioactive material among the ones to be confined, but tritium is not their driver of the risks for workers and for members of the public. Nevertheless, the tritium quantities and risks from fusion facilities create specificities for a specific standard (e.g. in fusion facilities, tritium is the driver of routine and accident consequences). Therefore, the scope of this document does not cover the other facilities involved in tritium releases (ISO 17873, ISO 16647 and ISO 26802), even though these other facilities create tritium releases (e.g. non-reactor fission facilities, tritium laboratories, tritium removal facilities from fission plants, tritium defence facilities).

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 provides the general requirements associated with the specific characteristics of high efficiency particulate air (HEPA) filters used in nuclear facilities. This document provides the manufacturer with general requirements for the performance, design, construction, acceptance testing and quality assurance for HEPA filters used in nuclear facilities (for qualification and production tests). All types of HEPA filter used in such applications are covered, from the large size HEPA filters in exhaust HVAC systems to small size low flow rate cylindrical HEPA filters for glove boxes. The design, fabrication, inspection and testing, certificates with regards to their expected performances are mentioned. This document does not provide the specific conditions against which the nuclear filters are designed, tested and qualified. This document applies only to the filters used for nuclear heating ventilation air conditioning (HVAC) or control rooms habitability applications or applications related to the exposure to radioactive ionizing radiations (e.g. medical or radioactive aerosols applications) in the severe conditions (e.g. fire, high radioactive challenge). Filter housing qualification is not part of this document.

This document presents general principles for preparedness to conduct individual contamination screening, triage, monitoring and assessing radiation doses received by people exposed during and/or in the aftermath of a nuclear or major radiological incident. The document mainly focuses on the early response phase, which requires rapid actions to be undertaken for achieving the goals in support of, and according to, national or international guidelines on emergency response. It addresses general requirements for — members of the public, this includes adults, vulnerable populations (such as children and pregnant women) and people with special needs (such as the elderly and disabled), and — emergency workers. This document provides general procedures for screening, triage and monitoring these two categories of people. It deals with individual monitoring for potential external contamination, internal and external exposures and dose assessment. It also gives principles for organizing and managing a population screening centre and for registering and reporting the results of individual monitoring. This document is applicable to most exposure situations following a nuclear or major radiological incident affecting a large number of people, including: — significant release of radioactive materials (e.g. from a facility or nuclear power plant, during transportation); — radiological dispersal device (RDD); — improvised nuclear device (IND); — nuclear weapon. Radiological incidents for which there is no release of radioactive material in the environment but only external exposures (e.g. linked to a Radiation Exposure Device (RED)) are outside the scope of this document[1]. However, some information given by this document may be of interest for this type of event. The aim of the document is to ensure that the appropriate parties are prepared in advance. This document advises how to obtain and collect data quickly and accurately in order to inform decision makers. It does not specify the parties or individuals who are responsible for undertaking the actions. This document is intended to give guidance to those in charge of monitoring and assessing doses received by populations in emergency exposure situations involving a large number of people potentially subject to internal/external contamination (and subsequent radiation doses). It can also serve as guidance to regulatory bodies. [1] Incidents resulting from RED exposure are excluded from consideration in this document because they do not result in contamination that would be detected by a portal monitor or handheld device. Identification of victims with only potential external exposure are determined by means such as evaluation of clinical signs and symptoms, biodosimetry, EPR, etc.

The objective of this document is to promote the harmonization of data and information reporting formats in order to provide the basis for the evaluation of occupational exposure with a view to allow for benchmarking capacity at the user level, technical review level, country level and global level (such as UNSCEAR) database or register on occupational exposure. Activity sectors and occupations (where employees are classified as occupationally exposed workers) that is included in this database or register as well as dose types and different values of interest concerning occupational exposure are described as follows. A typical national dose register (NDR): — contains personal, employment, and dosimetric data of occupational employment and wage statistics (OEWs) in the country. — assists national authorities in controlling and safekeeping of the occupational doses and to allow statistical evaluations (e.g., dose trends to answer requests from regulators and others). — assists in regulatory control by notifying regulatory authorities of overexposures within their jurisdiction and the licensee in their respective facility. — contributes to health research and to the scientific knowledge on risks from occupational exposure to ionizing radiation. — provides dose histories to individual workers and organizations for work planning and for compensation and litigation cases. All information provided by the NDR, including dose histories, may be subject to confidentiality requirements. This document is aimed at national dose registries but may be also applicable to dosimetry services that provide data to national dose registries. NOTE Such a database or register on occupational radiation dose for different sectors will, among other reasons, allow to prepare the data necessary for more global surveys, such as those undertaken by the UNSCEAR and other databases such as IAEA’s Information System on Occupational Exposure in Medicine, Industry and Research (ISEMIR), Information System on Occupational Exposure (ISOE) and the European Platform for Occupational Radiation Exposure (ESOREX‑Platform). Presently, as the formats are different, the international description of national statistics is often incomplete or inaccurate, and in the end, the comparison of data is not established yet in many countries. This standard defines a common and easily shared format to collect reliable, traceable and directly comparable data on individual and collective exposure in activity sectors and occupations as defined in a common way. This document addresses: a) a common list of activity sectors and occupations, and b) a common and easily shared format about dose types and different values of interest concerning occupational exposure in order to collect consistent and directly comparable data on individual and collective exposure.

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 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 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 specifies the calibration methods under laboratory conditions for dosemeters used for environmental and area monitoring of X and gamma-rays with respect to the operational quantities of the International Commission on Radiation Units and Measurements (ICRU)[1]. This document extends the dose rate range of ISO 4037-1 below 1,0 µSv·h−1. The specific uncertainty components are described for these calibration methods. This document also specifies the method for routine checking of active area dosemeters. Routine checking is not a calibration, nor does it replace a calibration, but it is a simple and effective method to routinely verify that the performance of the equipment is continuously maintained after calibration and that the calibration is still valid. This document does not deal with the special requirements for the calibration of spectrometer-based environmental dosemeters and passive dosemeters.

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 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 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 sets forth performance-based criteria and recommendations for the design and use of systems for sampling of airborne radioactive materials in the effluent air from the ducts and stacks of nuclear facilities. The requirements and recommendations of this document are aimed at sampling that is conducted for regulatory compliance and system control. If existing air-sampling systems are not designed to the performance requirements and recommendations of this document, an evaluation of the performance of the system is advised. If deficiencies are discovered, a determination of whether or not a retrofit is needed and practicable is recommended. It can be impossible to meet the requirements of this document in all conditions with a sampling system designed for normal operations only. Under off-normal conditions, the criteria or recommendations of this document still apply. However, for accident conditions, special accident air sampling systems or measurements can be used. This document does not address outdoor air sampling, radon measurements, or the surveillance of airborne radioactive substances in the workplace of nuclear facilities. NOTE Reference [1] addresses the instrumentation that is frequently used in nuclear air monitoring. Reference [5] addresses air sampling in the workplace of nuclear facilities. References [6] and [7] describe the performance characteristics of air monitors.

These international guidelines are based on the assumption that monitoring of environmental components (atmosphere, water, soil and biota) as well as food quality is performed to ensure the protection of human health[5][7][8][9][10][11][12]. The guidelines constitute a basis for the setting of national regulations, standards, and inter alia, for monitoring air, water and food in support of public health, specifically to protect the public from ionizing radiation. This document provides: — guidance to collect data needed for the assessment of human exposure to radionuclides naturally present or discharged by anthropogenic activities in the different environmental compartments (atmosphere, waters, soils, biota) and food; — guidance on the environmental characterization needed for the prospective and/or retrospective dose assessment methods of public exposure; — guidance that addresses actions appropriate for an event involving uncontrolled releases of gamma-emitters (e.g. nuclear power reactor emergencies) and also events that would involve beta- or alpha-emitters would require additional consideration of the pathways, instrumentation, laboratory analysis, operational intervention levels, protective actions, etc., appropriate to their release; — guidance for staff in nuclear installations responsible for the preparation of radiological assessments in support of permit or authorization applications and National Authorities’ officers in charge of the assessment of doses to the public for the purposes of determining gaseous or liquid effluent radioactive discharge authorizations; — information to the public on the parameters used to conduct a dose assessment for any exposure situations to a representative person/population. It is important that the dose assessment process be transparent, and that assumptions are clearly understood by stakeholders who can participate in, for example, the selection of habits of the representative person to be considered. This document refers to various published ISO documents. When appropriate, this document also refers to national standards or other publicly available documents.

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 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 describes a test method to determine the activity concentration of atmospheric tritium by trapping tritium in air by bubbling through a water solution. Atmospheric tritium activity concentration levels are expressed in becquerel per cubic metre (Bq∙m-3). 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 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 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. It specifies minimum requirements for proficiency testing applicable to dosimetry laboratories that have dedicated facilities for in vivo monitoring and where accreditation is required as part of providing the service. It also provides general requirements for proficiency testing that may include a larger group of non-accredited laboratories that may perform measurements as part of worker surveillance or in response to an emergency. 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 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 (Annex B). — 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. The activity concentration is expressed in becquerel per cubic metre (Bq∙m-3). 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. The test method is used in the context of a quality assurance management system (ISO/IEC 17025[2]). 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[3][4]. This document does not address to sampling of radionuclides bound to aerosol particles in the effluent air of nuclear facilities [see ISO 2889:2021][5]. The procedures described here facilitate the sampling of aerosol bound radionuclides. It is supposed to conform to the national and international requirements for monitoring programmes safety standards of IAEA[6]. The characteristics of the sampling location (coordinates, type of vegetation, obstacles) need to be documented prior to commencing the monitoring. The guidelines of the World Meteorology Organization (WMO) include the criteria for representative measurements of temperature, wind-speed, wind direction, humidity and precipitation for all the weather stations in the world[7].

The scope of ISO 16659 series is to provide different test methods aiming at assessing the efficiency of radioactive iodine traps in ventilation systems of nuclear facilities. The ISO 16659 series deals with iodine traps containing a solid sorbent — mainly activated and impregnated charcoal, the most common solid iodine sorbents used in the ventilation systems of nuclear facilities — as well as other sorbents for special conditions (e.g. high temperature zeolites). The scope of this document is to provide general and common requirements for the different test methods for industrial nuclear facilities. The different methods will be described in other specific parts of ISO 16659 series. Nuclear medicine applications are excluded from the scope of ISO 16659 series. In principle, ISO 16659 series is used mainly for filtering radioactive iodine, but other radioactive gases can also be trapped together with iodine. In such a case, some specificity may have to be adapted for these other radioactive gases in specific parts of ISO 16659 series. This document describes the main general requirements in order to check in situ the efficiency of the iodine traps, according to test conditions that are proposed to be as reproducible as possible.

A water phantom is used to ensure the accurate measurement of absorbed dose delivered by a radiation therapy machine as well as standardizing the dose distribution produced by the radiation therapy device. This document describes a detailed procedure for the construction and calibration of a polystyrene phantom and the results of its use in measuring the absorbed dose profile around the mechanical centre of a radiosurgery medical device, the full width at half maximum (FWHM) of the field and the physical penumbra at the mechanical centre, as well as the associated uncertainties. According to IAEA TRS-483 document, the most common design recommended in Gamma Knife® system is a hemisphere atop a water filled or compact polystyrene cylinder, and when using a polystyrene phantom, the measurement depth of the absorbed dose to water is reported to be the centre of the hemisphere with the radius of 8 cm. This document mainly describes the procedure for measuring the absorbed dose distribution around the mechanical centre of Gamma Knife® and obtaining the FWHM and penumbra from it. The developed phantom is made of polystyrene and has a hemispherical shape in accordance with the design suggested in IAEA TRS-483. This type of phantom is specific and adapted only for the Gamma Knife® radiosurgery facilities (PerfexionTM and IconTM models) and does not apply to general dosimetry protocols in radiotherapy facilities that use a small radiation field to treat a disease such as LINAC or Cyberknife. Considering that the type of medical device corresponds to treatment using external beam radiotherapy following small static fields, this technical report follows the recommendations published in the IAEA TRS‑483.

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.). For simplification, the term “soil” used in this document covers the set of elements mentioned above. This document is addressed to the people responsible for determining the radioactivity present in soil for the purpose of radiation protection. It is applicable to soil from gardens, 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 range of the testing performed. When a laboratory does not undertake one or more of the activities covered by this document, such as planning, sampling, test or calibration, the corresponding requirements do not apply. NOTE The term “laboratory” is applicable to all identified entities (individuals, organizations, etc.) performing planning, sampling, test and calibration.

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 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 0,06 MeV to 1,1 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[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 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 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 and 3,6 MeV maximum beta energy corresponding to 0,06 MeV to 1,1 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 Report 56[3], 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 beta reference 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 describes procedures for calibrating and determining the response of dosemeters and dose-rate meters in terms of the International Commission on Radiation Units and Measurements (ICRU) operational quantities for radiation protection purposes. However, as noted in ICRU 56[2], 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 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 ISO/IEC Guide 98-3:2008/Suppl.1 in ISO 11929-2, 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 ISO 11929-1:2019, Annex A the special case of repeated counting measurements with random influences and in ISO 11929-1:2019, Annex B, measurements with linear analogous ratemeters are covered. ISO 11929-2 extends ISO 11929-1 to the evaluation of measurement uncertainties according to the ISO/IEC Guide 98-3:2008/Suppl.1. ISO 11929-2 also presents some explanatory notes regarding general aspects of counting measurements and 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, alpha- and gamma-spectrometric measurements. Further, it provides some advice how to deal with correlations and covariances. 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. The ISO 11929 (all parts) also applies analogously to other measurements of any kind if a similar model of the evaluation is involved. Further practical examples can be found in other International Standards, for example, see References [1 to 20]. NOTE A code system, named UncertRadio, is available allowing for calculations according to ISO 11929-1 to ISO 11929-3. UncertRadio[40][41] 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 that 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’[44] 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, respectively. Covariances/correlations of input quantities are included. Applying these two functions within iterations for decision threshold and the detection limit calculations simplifies the programming effort significantly. It is also possible to implement this document in a spreadsheet containing a Monte Carlo add-in or into other commercial mathematics software.

This document presents the methods and provisions for sampling tritium and carbon‑14 in the gaseous effluents generated by nuclear facilities during operation and decommissioning. Specifically included are sample withdrawal location, extraction, transport flow measurement, and collection for later analysis. This document doesn’t address to real time measurements of tritium activity and carbon-14 activity in the effluent air of stacks and ducts. Information about real time measurements can be found in ISO 2889:2021, Annex H. Sample processing, analysis and calculations of tritium and carbon‑14 emissions will be addressed in future parts of ISO 20041.

This document addresses aspects of management of solid biomedical radioactive waste from its generation in nuclear medicine facilities to final clearance and disposal, as well as the manner to establish an effective program for biomedical radioactive waste management. Liquid and gaseous wastes are excluded from the scope of the document, but solid waste includes spent and surplus solutions of radionuclides contained in vials, tubes or syringes. Therefore, this document should be useful for any nuclear medicine facilities dealing with in vivo medical applications of radionuclides and consequently with the waste associated with such applications. This document provides a list of the main radionuclides used in nuclear medicine facilities and their main physical characteristics, as well as the guidance to write a radioactive waste management program for their sorting, collection, packaging and labelling, radioactivity surveys and decay storage, clearance levels, and transportation, if necessary, until their ultimate disposal or discharge. This document may also be useful as guidance for regulatory bodies.

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. Liquid standards that are intended to be used for preparing extended standards by the laboratories are also within the scope of this document. Reference materials (RMs) produced in accordance with ISO 17034 are out of scope of this document.

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. NOTE This document is directed at all stakeholders who are involved: IMSs, accreditation or regulatory bodies, and users of the particular dosimetry (the user is meant as the entity which assigns the dosimetry system to the radiation worker and records the assigned dose.) 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. The links between ISO 21909-1 and ISO 21909-2 are described in Annex E. This document only addresses neutron personal monitoring and not criticality accident conditions.

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 gives information for extremity dosimetry in the Annex C, based on recommendations given by ICRU Report 66. This document addresses only neutron personal monitoring and not criticality accident conditions. The links between this document and ISO 21909-2 are given in Annex A.