ISO/TC 85 - Nuclear energy, nuclear technologies, and radiological protection

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 guidance in the preparation, verification, and validation of group-averaged neutron and gamma-ray cross sections for the energy range and materials of importance in radiation protection and shielding calculations for nuclear reactors[1], see also Annex A. [1] This edition is based on ANSI/ANS-6.1.2-2013[1].

1.1 This is a practice for using a radiochromic optical waveguide dosimetry system to measure absorbed dose in materials irradiated by photons and high energy electrons in terms of absorbed dose to water. The radiochromic optical waveguide dosimetry system is generally used as a routine dosimetry system. 1.2 The optical waveguide dosimeter is classified as a Type II dosimeter on the basis of the complex effect of influence quantities (see ISO/ASTM Practice 52628). 1.3 This document is one of a set of standards that provides recommendations for properly implementing dosimetry in radiation processing, and describes a means of achieving compliance with the requirements of ISO/ASTM 52628 for an optical waveguide dosimetry system. It is intended to be read in conjunction with ISO/ASTM Practice 52628. 1.4 This practice applies to radiochromic optical waveguide dosimeters that can be used within part or all of the specified ranges as follows: 1.4.1 The absorbed dose range is from 1 Gy to 20 000 Gy. 1.4.2 The absorbed dose rate is from 0.001 Gy/s to 1000 Gy/s. 1.4.3 The radiation photon energy range is from 1 MeV to 10 MeV. 1.4.4 The radiation electron energy range is from 3 MeV to 25 MeV. 1.4.5 The irradiation temperature range is from –78 °C to +60 °C. 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.7 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

This document provides the basis for calculating the decay heat power of non-recycled nuclear fuel of light water reactors. For this purpose the following components are considered: — the contribution of the fission products from nuclear fission; — the contribution of the actinides; — the contribution of isotopes resulting from neutron capture in fission products. This document applies to light water reactors (pressurized water and boiling water reactors) loaded with a nuclear fuel mixture consisting of 235U and 238U. Application of the fission product contribution to decay heat developed using this document to other thermal reactor designs, including heavy water reactors, is permissible provided that the other contributions from actinides and neutron capture are determined for the specific reactor type. Its application to recycled nuclear fuel, like mixed-oxide or reprocessed uranium, is not permissible. The calculation procedures apply to decay heat periods from 0 s to 109 s.

This document is applicable to the determination of gadolinium as Gd2O3 in powder blends and sintered pellets of Gd2O3 + UO2 and ((U, Gd) O2) from mass fraction 10 g/kg to 100 g/kg (i.e. 1 % to 10 %), using a suitable ICP-AES instrument. This methodology is capable of demonstrating compliance with agreed upon fuel specifications and associated data quality objectives provided the user has performed qualification measurements under their established measurement control program to demonstrate that measurement uncertainty requirements will be met with the desired level of confidence at the specification.

This document 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.

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

This document specifies a method for the determination of uranium concentrations in nitric acid or TBP-DILUANT (for example TBP-kerosene) solutions coming from the nuclear fuel cycle. The method is applicable — for process control of solutions, free of suspension, which contain between 10 g/l to 300 g/l uranium, and — for high accuracy purposes (Safeguards) to nitric acid solutions, free of suspension, which contain between 100 g/l and 220 g/l uranium. Having — the content of neptunium and plutonium impurities in the solution less than 1 % of the uranium content. — the content of neutron poisons (gadolinium, erbium) less than 1 % of the uranium content to ensure the absence of significant interferences at the level of required precision, for high accuracy purposes. The method is applicable to solid samples as well, provided that they can be fully dissolved in nitric acid.

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.

This document specifies an analytical method for determining heavy water isotopic purity by Fourier transform infrared spectroscopy (FTIR). It is applicable to the determination of the whole range of heavy water concentration. The method is devoted to process controls at the different steps of the process systems in heavy water reactor power plant or any other related areas. The method can be applied for heavy water isotopic purity measurements in a heavy water reactor power plant or research reactor, heavy water production factory and heavy water related areas.

This document addresses the measurement methods, procedures and uncertainty estimation for the measurement, using a personal dosimeter, of the effective dose to the caregiver in the vicinity of the patient treated with radioiodine to ablate the thyroid. The general requirements for the patient and caregiver and a guidance (see Annex A) for designated expert on instructing caregivers of discharged patients is considered to effectively measure the effective dose to the caregiver in the vicinity of the patient.

This document describes radon-222 integrated measurement techniques with passive sampling. It gives indications for determining the average activity concentration of the radon-222 in the air from measurements based on easy-to-use and low-cost passive sampling, and the conditions of use for the sensors. This document covers samples taken without interruption over periods varying from a few days to one year. This measurement method is applicable to air samples with radon activity concentrations greater than 5 Bq/m3.

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

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

These international guidelines are based on the assumption that monitoring of environmental components (atmosphere, water, soil and biota) as well as food quality ensure the protection of human health[2][4][5][6][7][8]. The guidelines constitute a basis for the setting of national regulations and standards, 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, biological components) and food; — guidance on the environmental characterization needed for the prospective and/or retrospective dose assessment methods of public exposure; — 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 for 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. Generic mathematical models used for the assessment of radiological human exposure are presented to identify the parameters to monitor, in order to select, from the set of measurement results, the "best estimates" of these parameter values. More complex models are often used that require the knowledge of supplementary parameters. The reference and limit values are not included in this document.

This document specifies requirements for the ice plug technique with liquid nitrogen or dry ice as refrigerant (cryogenic medium) on metal pipes of nuclear power plants. The freezing liquid can be water or water mixture (e.g. boric acid mixture). This document specifies technical requirements of ice plug generation, formation judgment and removal, measures before, during and after ice plugging and requirements for personnel and non-destructive testing. The application of the ice plug isolation technique is principally not allowed on cladded pipes or pipes with internal coatings. The application for pressure test is not in the scope of this document and will be qualified separately.

This document contains the terms, definitions, notes to entry and examples corresponding to the frequently used concepts which apply to diagnostic and therapeutic nuclear medicine. It comprises the minimum essential information for each nuclear medicine concept represented by a single term. It provides the reader with the information required to approach this multidisciplinary speciality, such as medical, radiopharmacy and medical physics point of view. It is intended to facilitate communication and promote common understanding.

This document provides the following: — specifications for cylinders for the transport of uranium hexafluoride (UF6) to provide compatibility among different users, — description of cylinder designs, but is not intended to develop new designs, — fabrication requirements for the procurement of new cylinders designed for the transport of 0,1 kg or more of uranium hexafluoride, — fabrication requirements for the procurement of new valve protections, valves and plugs, and — requirements for cylinders and valve protections in service.

This document specifies the basic requirements of thermal insulation design of reactor coolant system (RCS) equipment and piping. Among thermal insulation of various RCS equipment and piping, the following two kinds of thermal insulations are described in detailed based on common design logic and requirements: — thermal insulation of reactor pressure vessel (RPV); — thermal insulation of RCS piping and other equipment. This document is valid for two types of thermal insulation: — metallic thermal insulation; — non-metallic thermal insulation. This document mainly applies to nuclear power plants with pressurized water reactor (PWR). For other reactor types, this document can be taken as reference.

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

This document specifies 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].

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 series). The test method is not suitable for use of radiochemicals if the radionuclide emit low energy gamma rays or beta particles that are readily attenuated in the surface.

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

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

This document specifies methods and procedures for characterizing the responses of devices used for the determination of ambient dose equivalent for the evaluation of exposure to cosmic radiation in civilian aircraft. The methods and procedures are intended to be understood as minimum requirements.

This document specifies the conceptual basis for the determination of ambient dose equivalent for the evaluation of exposure to cosmic radiation in civilian aircraft and for the calibration of instruments used for that purpose.

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

This practice covers dosimetric procedures to be followed in installation qualification, operational qualification and performance qualification (IQ, OQ, PQ), and routine processing at electron beam facilities to ensure that the product has been treated with an acceptable range of absorbed doses. Other procedures related to IQ, OQ, PQ, and routine product processing that may influence absorbed dose in the product are also discussed. The electron beam energy range covered in this practice is between 80 and 300 keV, generally referred to as low energy. Dosimetry is only one component of a total quality assurance program for an irradiation facility. Other measures may be required for specific applications such as medical device sterilization and food preservation. Other specific ISO and ASTM standards exist for the irradiation of food and the radiation sterilization of health care products. For the radiation sterilization of health care products, see ISO 11137-1. In those areas covered by ISO 11137-1, that standard takes precedence. For food irradiation, see ISO 14470. Information about effective or regulatory dose limits for food products is not within the scope of this practice (see ASTM F1355 and F1356). This document is one of a set of standards that provides recommendations for properly implementing dosimetry in radiation processing, and describes a means of achieving compliance with the requirements of ISO/ASTM 52628. It is intended to be read in conjunction with ISO/ASTM 52628. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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

This document gives guidelines for pre-service inspection (PSI) and in-service inspections (ISI) for reactor coolant circuit components of light water reactors and their installations as direct or remote visual testing in the form of a — general visual testing (overview), or — selective visual testing (specific properties). This document is also applicable to other components of nuclear installations. The requirements in this document focuses on remote (mechanized) visual testing, but also specifies global requirements for direct visual testing. For specific requirements for direct visual testing of welds see ISO 17637. This document is not applicable to tests in respect to the general state that are carried out in conjunction with pressure and leak tests and regular plant inspections. This document specifies test methods that allow deviations from the expected state to be recognised, requirements for the equipment technology and test personnel, the preparation and performance of the testing as well as the recording. NOTE Data concerning the test section, test extent, inspection period, inspection interval and evaluation of indications is defined in the applicable national nuclear safety standards.

This document gives guidelines for pre-service inspection (PSI) and in-service inspections (ISI) by eddy current tests on non-ferromagnetic steam generator heating tubes of light water reactors, whereby the test is carried out using mechanised test equipment outwards from the tube inner side. An in-service eddy current test of steam generator heating tube plugs as a component of the primary circuit is not an object of this document. Owing to the different embodiments of steam generator heating tube plugs, the use of specially adapted test systems to be qualified is necessary. Test systems for the localisation of inhomogeneities (surface) and requirements for test personnel, test devices, the preparation of test and device systems, the implementation of the testing as well as the recording are defined. NOTE Data concerning the test section, test extent, inspection period, inspection interval and evaluation of indications is defined in the nuclear safety standards. It is recommended that the technical specifications are based on experience on U-tube bends with even bend radius (similar to the S/KWU design). To test other kind of tube bends (e.g. U-tube bends with two 90° bends) further qualifications will be provided.

This document gives guidelines for pre-service-inspections (PSI) and in-service inspections (ISI) with mechanized ultrasonic test (UT) devices on components of the reactor coolant circuit of light water reactors. This document is also applicable on other components of nuclear installations. Mechanized ultrasonic inspections are carried out in order to enable an evaluation in case of — fault indications (e.g. on austenitic weld seams or complex geometry), — indications due to geometry (e.g. in case of root concavity), — complex geometries (e.g. fitting weld seams), or — if a reduction in the radiation exposure of the test personnel can be attained in this way. Ultrasonic test methods are defined for the validation of discontinuities (volume or surface open), requirements for the ultrasonic test equipment, for the preparation of test and device systems, for the implementation of the test and for the recording. This document is applicable for the detection of indications by UT using normal-beam probes and angle-beam probes both in contact technique. It is to be used for UT examination on ferritic and austenitic welds and base material as search techniques and for comparison with acceptance criteria by the national referencing nuclear safety standards. Immersion technique and techniques for sizing are not in the scope of this document and are independent qualified. NOTE Data concerning the test section, test extent, inspection period, inspection interval and evaluation of indications is defined in the applicable national nuclear safety standards. Unless otherwise specified in national nuclear safety standards the minimum requirements of this document are applicable. This document does not define: — extent of examination and scanning plans; — acceptance criteria; — UT techniques for dissimilar metal welds and for sizing (have to be qualified separately); — immersion techniques; — time-of-flight diffraction technique (TOFD). It is recommended that UT examinations are nearly related to the component, the type and size of defects to be considered and are reviewed in specific national inspection qualifications.

This document gives guidelines for in-service system pressure tests of the reactor coolant circuit of light water reactors. This document specifies the test technique, the requirements for measuring equipment and additional devices, the preparation and performance of the test as well as the recording and documentation, for the purpose to ensure the reliability and comparability of tests. NOTE Data on (test) pressure, (test) temperature, scope of testing, rates of change of pressure and temperature, test schedule and inspection intervals can be obtained from the applicable national nuclear codes.

This document gives guidelines for pre-service inspections (PSI) and in-service inspections (ISI) of the surfaces using the magnetic particle testing and penetrant testing on components of the reactor coolant circuit of light water reactors. This document is also applicable to other components of nuclear installations. Test systems for the localisation of surface inhomogeneities and requirements for test personnel, test devices, test media, accessories as well as optical auxiliaries, the preparation and implementation of the test as well as the recording are defined. NOTE 1 Data concerning the test section, test extent, inspection period, inspection interval and evaluation of indications are defined in the applicable national nuclear safety standards. NOTE 2 In general, this document is in accordance with ISO 3452 and ISO 9934 series. This document provides details to be considered in the standard test procedure (see Annex A).

The use of a continuous air monitor (CAM) is mainly motivated by the need to be alerted quickly and in the most accurate way possible with an acceptable false alarm rate when significant activity concentration value is exceeded, in order to take appropriate measures to reduce exposure of those involved. The performance of this CAM does not only depend on the metrological aspect characterized by the decision threshold, the limit of detection and the measurement uncertainties but also on its dynamic capacity characterized by its response time as well as on the minimum detectable activity concentration corresponding to an acceptable false alarm rate. The ideal performance is to have a minimum detectable activity concentration as low as possible associated with a very short response time, but unfortunately these two criteria are in opposition. It is therefore important that the CAM and the choice of the adjustment parameters and the alarm levels be in line with the radiation protection objectives. This document describes — the dynamic behaviour and the determination of the response time, — the determination of the characteristic limits (decision threshold, detection limit, limits of the coverage interval), and — a possible way to determine the minimum detectable activity concentration and the alarms setup. Finally the annexes of this document show actual examples of CAM data which illustrate how to quantify the CAM performance by determining the response time, the characteristics limits, the minimum detectable activity concentration and the alarms setup.

The use of a continuous air monitor (CAM) is mainly motivated by the need to be alerted quickly and in the most accurate way possible with an acceptable false alarm rate when a significant activity concentration value is exceeded, in order to take appropriate measures to reduce exposure of those involved. The performance of this CAM does not only depend on the metrological aspect characterized by the decision threshold, the limit of detection and the measurement uncertainties but also on its dynamic capacity characterized by its response time as well as on the minimum detectable activity concentration corresponding to an acceptable false alarm rate. The ideal performance is to have a minimum detectable activity concentration as low as possible associated with a very short response time, but unfortunately these two criteria are in opposition. It is therefore important that the CAM and the choice of the adjustment parameters and the alarm levels be in line with the radiation protection objectives. The knowledge of a few factors is needed to interpret the response of a CAM and to select the appropriate CAM type and its operating parameters. Among those factors, it is important to know the half-lives of the radionuclides involved, in order to select the appropriate detection system and its associated model of evaluation. CAM using filter media accumulation sampling techniques are usually of two types: a) fixed filter; b) moving filter. This document first describes the theory of operation of each CAM type i.e.: — the different models of evaluation considering short or long radionuclides half-lives values, — the dynamic behaviour and the determination of the response time. In most case, CAM is used when radionuclides with important radiotoxicities are involved (small value of ALI). Those radionuclides have usually long half-life values. Then the determination of the characteristic limits (decision threshold, detection limit, limits of the coverage interval) of a CAM is described by the use of long half-life models of evaluation. Finally, a possible way to determine the minimum detectable activity concentration and the alarms setup is pointed out. The annexes of this document show actual examples of CAM data which illustrate how to quantify the CAM performance by determining the response time, the characteristics limits, the minimum detectable activity concentration and the alarms setup.

This document contains the terms, definitions, notes to entry and examples corresponding to the basic concepts of the nuclear energy, nuclear technologies, and radiological protection subject fields. It provides the minimum essential information for each cross-cutting concept represented by a single term. NOTE A full understanding of concepts goes with a background knowledge of nuclear energy, nuclear technologies, and radiological protection. It is intended to facilitate communication and promote common understanding.

This practice describes the basic requirements that apply when making absorbed dose measurements in accordance with the ASTM E61 series of dosimetry standards. In addition, it provides guidance on the selection of dosimetry systems and directs the user to other standards that provide specific information on individual dosimetry systems, calibration methods, uncertainty estimation and radiation processing applications. This practice applies to dosimetry for radiation processing applications using electrons or photons (gamma- or X-radiation). This practice addresses the minimum requirements of a measurement management system, but does not include general quality system requirements. This practice does not address personnel dosimetry or medical dosimetry. This practice does not apply to primary standard dosimetry systems. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use.

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

This practice covers the preparation and use of semiadiabatic calorimetric dosimetry systems for measurement of absorbed dose and for calibration of routine dosimetry systems when irradiated with electrons for radiation processing applications. The calorimeters are either transported by a conveyor past a scanned electron beam or are stationary in a broadened beam. This document is one of a set of standards that provides recommendations for properly implementing dosimetry in radiation processing, and describes a means of achieving compliance with the requirements of ISO/ASTM Practice 52628 for a calorimetric dosimetry system. It is intended to be read in conjunction with ISO/ASTM Practice 52628. The calorimeters described in this practice are classified as Type II dosimeters on the basis of the complex effect of influence quantities. See ISO/ASTM Practice 52628. This practice applies to electron beams in the energy range from 1.5 to 12 MeV. The absorbed dose range depends on the calorimetric absorbing material and the irradiation and measurement conditions. Minimum dose is approximately 100 Gy and maximum dose is approximately 50 kGy. The average absorbed-dose rate range shall generally be greater than 10 Gy·s-1. The temperature range for use of these calorimetric dosimetry systems depends on the thermal resistance of the calorimetric materials, on the calibration range of the temperature sensor, and on the sensitivity of the measurement device. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety, health, and environmental practices and determine the applicability of regulatory limitations prior to use. This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee.

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

This document describes radon-222 spot measurement methods. It gives indications for carrying out spot measurements, at the scale of a few minutes at a given place, of the radon activity concentration in open and confined atmospheres. This measurement method is intended for rapid assessment of the radon activity concentration in the air. The result cannot be extrapolated to an annual estimate of the radon activity concentration. This type of measurement is therefore not applicable for assessment of the annual exposure or for determining whether or not to mitigate citizen exposures to radon or radon decay products. The measurement method described is applicable to air samples with radon activity concentration greater than 50 Bq·m−3. NOTE For example, using an appropriate device, the radon activity concentration can be spot measured in the soil and at the interface of a material with the atmosphere (see also ISO 11665-7[8]).

This document describes continuous measurement methods for radon-222. It gives indications for continuous measuring of the temporal variations of radon activity concentration in open or confined atmospheres. This document is intended for assessing temporal changes in radon activity concentration in the environment, in public buildings, in homes and in work places, as a function of influence quantities such as ventilation and/or meteorological conditions. The measurement method described is applicable to air samples with radon activity concentration greater than 5 Bq/m3.

This document describes spot measurement methods for determining the activity concentration of short-lived radon-222 decay products in the air and for calculating the potential alpha energy concentration. This document gives indications for performing a spot measurement of the potential alpha energy concentration, after sampling at a given place for several minutes, and the conditions of use for the measuring devices. The measurement method described is applicable for a rapid assessment of the potential alpha energy concentration. The result obtained cannot be extrapolated to an annual estimate potential alpha energy concentration of short-lived radon-222 decay products. Thus, this type of measurement is not applicable for the assessment of annual exposure or for determining whether or not to mitigate citizen exposures to radon or radon decay products. This measurement method is applicable to air samples with potential alpha energy concentration greater than 5 nJ/m3. NOTE This document does not address the potential contribution of radon-220 decay products.

This document provides specific requirements and guidance on the nuclear criticality safety of waste containing fissile nuclides, generated during normal operations. This document is intended to be used along-side and in addition to ISO 1709. This document applies specifically to the nuclear criticality safety of solid nuclear wastes. It also applies to residual quantities of liquids and/or slurries which are either intimately associated with the solid nuclear waste materials or arise as a result of processing or handling the waste. This document does not apply to bulk or process liquids (including higher concentration process solutions) or irradiated or un-irradiated fuel elements. NOTE The term fuel element is sometimes applied to fuel assembly, fuel bundle, fuel cluster, fuel rod, fuel plate, etc. All these terms are based on one or more fuel elements. A cylindrical fuel rod (sometimes referred to as a fuel pin) for a light-water-reactor is an example of a fuel element. All stages of the waste life cycle are within the scope of the document. This document can also be applied to the transport of solid nuclear waste outside the boundaries of nuclear establishments.

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

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

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

This document covers trunnion systems used for tie-down, tilting and/or lifting of a package of radioactive material during transport operations. Aspects included are the design, manufacture, maintenance, inspection and management system. Regulations which can apply during handling operation in nuclear facilities are not addressed in document. This document does not supersede any of the requirements of international or national regulations, concerning trunnions used for lifting and tie-down.

This document specifies the dose assessment method when an RPLD is used for dosimetry audit in external high-energy X-ray beam radiotherapy. The dosimetry for electron beams and X-ray beams of stereotactic radiotherapy, gamma‑ray of brachytherapy is not included in this version. This document addresses RPLD handling, measurement method, conversion of measured value to dose, necessary correction coefficient, and the performance requirements for RPLD systems, including the reader.