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

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

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