27.120 - Nuclear energy engineering
ICS 27.120 Details
Nuclear energy engineering
Kerntechnik
Energie nucleaire
Jedrska tehnika
General Information
Frequently Asked Questions
ICS 27.120 is a classification code in the International Classification for Standards (ICS) system. It covers "Nuclear energy engineering". The ICS is a hierarchical classification system used to organize international, regional, and national standards, facilitating the search and identification of standards across different fields.
There are 1474 standards classified under ICS 27.120 (Nuclear energy engineering). These standards are published by international and regional standardization bodies including ISO, IEC, CEN, CENELEC, and ETSI.
The International Classification for Standards (ICS) is a hierarchical classification system maintained by ISO to organize standards and related documents. It uses a three-level structure with field (2 digits), group (3 digits), and sub-group (2 digits) codes. The ICS helps users find standards by subject area and enables statistical analysis of standards development activities.
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IEC 63048-2:2026 is applicable to MRCSs for ground surveillance of a nuclear facility or in a radiological environment. This document describes the mission, operating conditions, reliability requirements, functional requirements, operational requirements, and test requirements of MRCS for ground surveillance of nuclear facilities.
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IEC 63048-1:2026 is applicable to MRCSs for ground surveillance of a nuclear facility or in a radiological environment. This document describes the mission, operating conditions, reliability requirements, functional requirements, operational requirements, and test requirements of MRCS for ground surveillance of nuclear facilities.
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This document sets out the fundamental principles and procedure of determining the groundwater remediation targets for post in situ leach (also referred as in situ recovery) uranium mining. It also delineates the environmental investigation requirements for groundwater remediation, stakeholder identification and engagement, groundwater remediation technical options and their analysis of effectiveness, and considerations for setting and adjusting the groundwater remediation target values. This document is applicable to controlling and mitigating groundwater impact and conducting groundwater remediation for in situ leach uranium mining. It can also serve as a reference for contamination control, mitigation and remediation of groundwater for other in situ leach metal mining projects.
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This document specifies a method which applies to the preparation and validation of the standard materials generally called “large size spikes” with an uncertainty suitable for international nuclear safeguards used for measuring the content of plutonium and/or uranium by isotope dilution mass spectrometry.
This measurement methodology can be applied to input solutions of irradiated Magnox and light water reactor fuels (boiling water reactor or pressurized water reactor); in final products at spent-fuel reprocessing plants; in feed and products of mixed oxide of plutonium and uranium (MOX); and in uranium fuel fabrication
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This document specifies an analytical method for determining the neptunium concentration by spectrophotometry, with spectrophotometer implemented in hot cell or glove box allowing the analysis of high activity solutions, with a standard uncertainty, with coverage factor k = 1 of about 5 %, in nitric acid solutions after the dissolution of nuclear reactor irradiated fuels, at different steps of the process in a nuclear fuel reprocessing plant or in other nuclear facilities. The method is applicable to sample from the process containing a concentration of neptunium between 10 mg·l-1 and 400 mg·l-1 and uranium concentrations of up to 300 g·l-1.
- Standard16 pagesEnglish languagee-Library read for1 day
This document describes an analytical method for the determination of uranium in samples from pure product materials such as U metal, UO2, UO3, U3O8, uranyl nitrate hexahydrate and uranium hexafluoride from the nuclear fuel cycle. This procedure is sufficiently accurate and precise to be used for nuclear materials accountability. This method can be used directly for the analysis of most uranium and uranium oxide nuclear reactor fuels, either irradiated or un-irradiated, and of uranium nitrate product solutions. Fission products equivalent to up to 10 % burn-up of heavy atoms do not interfere, and other elements which could cause interference are not normally present in sufficient quantity to affect the result significantly. The method recommends that an aliquot of sample is weighed and that a mass titration is used, in order to obtain improved precision and accuracy. This does not preclude the use of alternative techniques which could give equivalent performance. The use of automatic device(s) in the performance of some critical steps of the method has some advantages, mainly in the case of routine analysis.
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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.
- Standard42 pagesEnglish languagee-Library read for1 day
This document describes an analytical method for the determination of uranium in samples from pure product materials such as U metal, UO2, UO3, uranyl nitrate hexahydrate, uranium hexafluoride and U3O8 from the nuclear fuel cycle. This procedure is sufficiently accurate and precise to be used for nuclear materials accountability. This method can be used directly for the analysis of most uranium and uranium oxide nuclear reactor fuels, either irradiated or un-irradiated, and of uranium nitrate product solutions. Fission products equivalent to up to 10 % burn-up of heavy atoms do not interfere, and other elements which could cause interference are not normally present in sufficient quantity to affect the result significantly. The method recommends that an aliquot of sample is weighed and that a mass titration is used, in order to obtain improved precision and accuracy. This does not preclude the use of alternative techniques which could give equivalent performance. The use of automatic device(s) in the performance of some critical steps of the method has some advantages, mainly in the case of routine analysis.
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This document specifies the requirements for the design and performance evaluation of the ECCS strainer, since the performance of the strainer is important for the maintenance of nuclear safety during DBA, DBE and DEC. This document is applicable to the design and performance evaluation of the ECCS strainer in PWR NPPs, particularly those with primary design philosophies that rely on ECCS pumps. Its relevance for some PWR designs could nevertheless need some adaptations. Moreover, although tailored to PWR NPPs, it can also serve as a valuable reference for other reactor types, providing insight into the design and effectiveness of filtration systems in various nuclear power applications. Additional information and insights on national practice can be found in References [1] [2] [3] [4].
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This document provides guidance for performing and validating the sequence of steady-state calculations leading to prediction in all types of operating commercial nuclear reactors, of the following: reaction-rate spatial distributions; reactivity; change of nuclide compositions with time. The document provides guidance for the selection of computational methods, criteria for verification and validation of calculation methods used by reactor core analysts, criteria for evaluation of accuracy and range of applicability of data and methods, and requirements for documentation of the preceding.
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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.
- Standard42 pagesEnglish languagee-Library read for1 day
This document specifies a method which applies to the preparation and validation of the standard materials generally called “large size spikes” with an uncertainty suitable for international nuclear safeguards used for measuring the content of plutonium and/or uranium by isotope dilution mass spectrometry.
This measurement methodology can be applied to input solutions of irradiated Magnox and light water reactor fuels (boiling water reactor or pressurized water reactor); in final products at spent-fuel reprocessing plants; in feed and products of mixed oxide of plutonium and uranium (MOX); and in uranium fuel fabrication
- Standard20 pagesEnglish languagee-Library read for1 day
This document describes an analytical method for the determination of uranium in samples from pure product materials such as U metal, UO2, UO3, uranyl nitrate hexahydrate, uranium hexafluoride and U3O8 from the nuclear fuel cycle. This procedure is sufficiently accurate and precise to be used for nuclear materials accountability. This method can be used directly for the analysis of most uranium and uranium oxide nuclear reactor fuels, either irradiated or un-irradiated, and of uranium nitrate product solutions. Fission products equivalent to up to 10 % burn-up of heavy atoms do not interfere, and other elements which could cause interference are not normally present in sufficient quantity to affect the result significantly. The method recommends that an aliquot of sample is weighed and that a mass titration is used, in order to obtain improved precision and accuracy. This does not preclude the use of alternative techniques which could give equivalent performance. The use of automatic device(s) in the performance of some critical steps of the method has some advantages, mainly in the case of routine analysis.
- Standard23 pagesEnglish languagee-Library read for1 day
This document specifies an analytical method for determining the neptunium concentration by spectrophotometry, with spectrophotometer implemented in hot cell or glove box allowing the analysis of high activity solutions, with a standard uncertainty, with coverage factor k = 1 of about 5 %, in nitric acid solutions after the dissolution of nuclear reactor irradiated fuels, at different steps of the process in a nuclear fuel reprocessing plant or in other nuclear facilities. The method is applicable to sample from the process containing a concentration of neptunium between 10 mg·l-1 and 400 mg·l-1 and uranium concentrations of up to 300 g·l-1.
- Standard16 pagesEnglish languagee-Library read for1 day
This document describes an analytical method for the determination of uranium in samples from pure product materials such as U metal, UO2, UO3, U3O8, uranyl nitrate hexahydrate and uranium hexafluoride from the nuclear fuel cycle. This procedure is sufficiently accurate and precise to be used for nuclear materials accountability. This method can be used directly for the analysis of most uranium and uranium oxide nuclear reactor fuels, either irradiated or un-irradiated, and of uranium nitrate product solutions. Fission products equivalent to up to 10 % burn-up of heavy atoms do not interfere, and other elements which could cause interference are not normally present in sufficient quantity to affect the result significantly. The method recommends that an aliquot of sample is weighed and that a mass titration is used, in order to obtain improved precision and accuracy. This does not preclude the use of alternative techniques which could give equivalent performance. The use of automatic device(s) in the performance of some critical steps of the method has some advantages, mainly in the case of routine analysis.
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This document specifies the methods and the requirements for the supersonic molecular beam injection (SMBI) fuelling technique used in experimental magnetic confinement fusion facilities. It outlines the SMBI system components, specific requirements, and inspection procedures to ensure the effective and controlled injection of plasma fuel into fusion devices such as international thermonuclear experimental reactor and demonstration power plant. This document applies to the formation of supersonic molecular beam (SMB) and usage of the SMBI technique on the fusion devices, including the specification of gas pressure adjustment, beam characteristics and injection rate requirements, as well as the procedures for verifying these parameters.
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This document specifies the performance and the functional characteristics of the low voltage static uninterruptible power supply (SUPS) systems in a nuclear power plant (NPP) and, when applicable, in nuclear facilities in general. An uninterruptible power supply (UPS) is an electrical equipment which draws electrical energy from a source, stores it, and maintains the supply in a specified form by means inside the equipment to output terminals. A SUPS has no rotating parts to perform its functions.
The specific design requirements for the components of the power supply system are covered by IEC standards and other standards listed in the normative references. Otherwise, specific component-level design requirements are outside the scope of this document.
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See the scope of IEC 61225:2025. Adoption is to be implemented without modification.
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This part of the IEC/IEEE 62582 series contains requirements for application of the other parts of IEC/IEEE 62582 related to specific methods for condition monitoring in electrical equipment important to safety of nuclear power plants. It also includes requirements which are common to all methods. The procedures defined in IEC/IEEE 62582 are intended for detailed condition monitoring.
IEC/IEEE 62582 specifies condition monitoring methods in sufficient detail to enhance the accuracy and repeatability, and provide standard formats for reporting the results. The methods specified are applicable to electrical equipment containing polymeric materials. Some methods are especially designed for the measurement of condition of a limited range of equipment whilst others can be applied to all types of equipment for which the polymeric parts are accessible.
Although the scope of IEC/IEEE 62582 is limited to the application of instrumentation and control systems important to safety, the condition monitoring methods can also be applicable to other components which include polymeric materials.
The different parts of IEC/IEEE 62582 are measurement standards, primarily for use in the management of ageing in initial qualification and after installation. For the technical background of condition monitoring methods, reference is made to other IEC standards, e.g. IEC 60544-5 [1]. Information on the role of condition monitoring in qualification of electrical equipment important to safety is found in IEC/IEEE 60780-323. General information on management of ageing can be found in IEC 62342 [5] and IEEE 1205 [6].
NOTE A simplified version of the procedures can be appropriate for preliminary assessment of the need for detailed measurements.
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IEC 62705:2022 gives requirements for the lifecycle management of radiation monitoring systems (RMS) and gives guidance on the application of existing IEC standards covering the design and qualification of systems and equipment. The purpose of this document is to lay down requirements for the lifecycle management of RMSs and give application guidance. This document is intended to be consistent with the latest versions of International Standards dealing with radiation monitors, sampling of radioactive materials, instruments calibration, hardware and software design, classification, and qualification. This document is applicable to RMSs installed in nuclear facilities intended for use during normal operation, anticipated operational occurrences (AOO), design basis accidents (DBA) and design extension conditions (DEC), including severe accidents (SA). This second edition cancels and replaces the first edition published in 2014. This edition includes the following significant technical changes with respect to the previous edition:
- modification of the title.
- to be consistent with the categorization of the accident condition.
- to update the references to new standards published since the first edition.
- to update the terms and definitions.
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This part of IEC/IEEE 62582 contains methods for condition monitoring of organic and polymeric materials in instrumentation and control systems using tensile elongation techniques in the detail necessary to produce accurate and reproducible measurements. This document includes the requirements for selection of samples, the measurement system and conditions, and the reporting of the measurement results.
The different parts of IEC/IEEE 62582 are measurement standards, primarily for use in the management of ageing in initial qualification and after installation. IEC/IEEE 62582-1 includes requirements for the application of the other parts of IEC/IEEE 62582 and some elements which are common to all methods. Information on the role of condition monitoring in qualification of equipment important to safety is found in IEC/IEEE 60780-323.
This document is applicable to non-energised equipment.
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IEC 62397:2022 describes the requirements for resistance temperature detectors (RTDs) suitable for applications in I&C systems important to safety of nuclear power plants. The requirements of RTDs include design, materials, manufacturing, testing, calibration, procurement, and inspection. RTDs used for safety applications in Nuclear Power Plants can be categorized into direct-immersed and thermowell-mounted RTDs.
This standard describes the requirements for the design, material selection, procurement, construction, and testing of resistance temperature detectors (RTDs) used in nuclear power plants (NPPs). These RTDs may be used in both the nuclear safety I&C systems and/or in the non-safety-related instrumentation systems.
This second edition cancels and replaces the first edition, published in 2007; it also cancels and replaces the first edition of IEC 61224:1993. This edition includes the following significant technical changes with respect to the previous edition.
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This part of IEC/IEEE 62582 specifies methods for condition monitoring of organic and polymeric materials in instrumentation and control systems using oxidation induction techniques in the detail necessary to produce accurate and reproducible measurements. It includes the requirements for sample preparation, the measurement system and conditions, and the reporting of the measurement results.
The different parts of IEC/IEEE 62582 are measurement standards, primarily for use in the management of ageing in initial qualification and after installation. IEC/IEEE 62582-1 includes requirements for the application of the other parts of the IEC/IEEE 62582 series and some elements which are common to all methods. Information on the role of condition monitoring in the qualification of equipment important to safety is found in IEC/IEEE 60780-323.
- Standard31 pagesEnglish languagee-Library read for1 day
This part of IEC/IEEE 62582 contains methods for condition monitoring of organic and polymeric materials in instrumentation and control systems using the indenter measurement technique in the detail necessary to produce accurate and reproducible measurements. It includes the requirements for the selection of samples, the measurement system and measurement conditions, and the reporting of the measurement results.
The different parts of IEC/IEEE 62582 are measurement standards, primarily for use in the management of ageing in initial qualification and after installation. IEC/IEEE 62582-1 includes requirements for the application of the other parts of the IEC/IEEE 62582 series and some elements which are common to all methods. Information on the role of condition monitoring in the qualification of equipment important to safety is found in IEC/IEEE 60780-323.
This document is intended for application to non-energised equipment.
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This document specifies a test method using radioactive methyl iodide (CH3131I) as a tracer to determine the in-situ decontamination factor of an iodine trap. An in-situ test allows to reach the global efficiency of the trap characterized by the sorbent efficiency but also by the implementation of the trap within the ventilation duct) while the intrinsic efficiency of a charcoal is characterized in a laboratory by ISO 18417[9] (or other national standards such as ASTM D3803[10]). This document provides general and common requirements for this method to assess the efficiency of an iodine trap, but also, the tools requirements, accuracy and the provisions needed to ensure safety of the workers, public and the environment during the test. This reproducible method can support nuclear facility operators as a reference method to compare the decontamination factor evaluated by this method to reference values (e.g. safety criteria, national legislation, etc.). Because of the use of a radioactive tracer, some precautions should be applied. Firstly, this method is usually used for ventilation systems with monitoring of gaseous iodine releases in environment in accordance with the national regulations. Secondly, this method is not used to determine the decontamination factor of iodine traps used in ventilation systems with air release in rooms with potential presence of workers (e.g. control room). For those rooms, a non-radioactive method is preferred. This document can apply to installations with low inventory of radioiodine equipped with iodine traps (e.g. small laboratories). In this case, some provisions can be adapted but always in accordance with the national regulations. Finally, this document mainly deals with iodine traps using impregnated activated carbon. However, this method can be used with some adaptations to other solid sorbent as inorganic sorbent (e.g. zeolite – aluminium and silica base usually doped with silver nitrate – or impregnated catalytic supports[11][12]).
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IEC 63506:2026 specifies the calibration method of prompt fission neutron logging tools. This document applies to the calibration of prompt fission neutron logging tools for uranium exploration and mining.
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IEC 61513:2026 provides requirements and recommendations for the overall I&C which may contain one or several of these technologies.
This document highlights also the need for complete and precise requirements, derived from the plant safety goals, as a pre-requisite for generating the comprehensive requirements for the overall I&C, and hence for the individual I&C systems important to safety.
This document introduces the concept of a safety lifecycle for the overall I&C including the I&C architecture, and a safety lifecycle for the individual systems. By this, it highlights the relations between the safety objectives of the NPP and the requirements for the architecture of the I&C systems important to safety, and the relations between the I&C architecture and the requirements of the individual systems important to safety.
Standards such as ISO/IEC/IEEE 15288 provide an overarching concept of system life cycle provisions covering product-related processes as well business development. The scope of IEC 61513 refers to safety aspects and their demonstration and significantly deepens the considerations of ISO/IEC/IEEE 15288 in this field.
The lifecycles illustrated in and followed by this document are not the only ones possible; other lifecycles can be followed, provided that the requirements stated in this document are satisfied.
This document applies to the I&C of new nuclear power plants as well as to I&C upgrading or back-fitting of existing plants.
This third edition cancels and replaces the second edition, published in 2011. This edition includes the following significant technical changes with respect to the previous edition:
- to align this document with the recent IAEA documents SSR-2/1 and SSG-39,
- to review the existing requirements and to update the terminology and definitions;
- to take account of, as far as possible, requirements associated with standards published or thoroughly revised since the second edition, IEC 62566, IEC 61226, IEC 62138, IEC 60987, IEC 63046 and IEC 63351;
- to incorporate the technical requirements related to I&C systems and equipment from IEC 61226:2020, Annex A.
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See the scope of IEC/IEEE 62582-1:2024. Adoption is to be implemented without modification.
- Standard24 pagesEnglish languagee-Library read for1 day
IEC 62397:2022 describes the requirements for resistance temperature detectors (RTDs) suitable for applications in I&C systems important to safety of nuclear power plants. The requirements of RTDs include design, materials, manufacturing, testing, calibration, procurement, and inspection. RTDs used for safety applications in Nuclear Power Plants can be categorized into direct-immersed and thermowell-mounted RTDs. This standard describes the requirements for the design, material selection, procurement, construction, and testing of resistance temperature detectors (RTDs) used in nuclear power plants (NPPs). These RTDs may be used in both the nuclear safety I&C systems and/or in the non-safety-related instrumentation systems. This second edition cancels and replaces the first edition, published in 2007; it also cancels and replaces the first edition of IEC 61224:1993. This edition includes the following significant technical changes with respect to the previous edition.
- Standard42 pagesEnglish languagee-Library read for1 day
IEC 62705:2022 gives requirements for the lifecycle management of radiation monitoring systems (RMS) and gives guidance on the application of existing IEC standards covering the design and qualification of systems and equipment. The purpose of this document is to lay down requirements for the lifecycle management of RMSs and give application guidance. This document is intended to be consistent with the latest versions of International Standards dealing with radiation monitors, sampling of radioactive materials, instruments calibration, hardware and software design, classification, and qualification. This document is applicable to RMSs installed in nuclear facilities intended for use during normal operation, anticipated operational occurrences (AOO), design basis accidents (DBA) and design extension conditions (DEC), including severe accidents (SA). This second edition cancels and replaces the first edition published in 2014. This edition includes the following significant technical changes with respect to the previous edition: - modification of the title. - to be consistent with the categorization of the accident condition. - to update the references to new standards published since the first edition. - to update the terms and definitions.
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See the scope of IEC/IEEE 62582-4:2022. Adoption is to be implemented without modification.
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See the scope of IEC/IEEE 62582-3:2024. Adoption is to be implemented without modification.
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See the scope of IEC/IEEE 62582-2:2022. Adoption is to be implemented without modification.
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ISO 16659 series provide different test methods aiming at assessing the performances of radioactive iodine traps in ventilation systems of nuclear facilities. This series deals with iodine traps with solid sorbent, mainly activated and impregnated charcoal, the most common solid sorbents used in ventilation systems of nuclear facilities, as well as other sorbents for special conditions (e.g. high temperature zeolites). ISO 16659-1 provides the general requirements to be applied for all methods of the series. The scope of this document is to provide general and generic requirements for the test method using cyclohexane (C6H12) as a tracer to determine the mechanical leakage rate of iodine trap. This reproducible method can support nuclear operators to compare the result with reference values given in safety reports. Unlike the method of radioactive methyl iodide described in ISO 16659-2, the cyclohexane field test method covered in this document does not directly give a decontamination factor for the iodine trap, but only the iodine trap performance information of an integrity test. Due to the use of the environmentally friendly test reagent of low-toxicity in the field tests, the method is mainly suitable for ventilation systems of those habitable spaces (e.g. main control rooms of nuclear power plants), and performance test of a single iodine trap before its delivery and acceptance. In addition, the method can also be used for iodine traps with activated carbon sampling canister (e.g. deep bed iodine trap Type III and Drawer iodine trap Type II).
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This part of IEC 62676 specifies the functions, performance, interfaces, environmental adaptability, test methods, performance evaluation and grading rules of real-time intelligent video analysis in surveillance systems.
This document applies to live and forensic, real-time intelligent video analysis devices and systems in video surveillance.
The document is centred on testing performance and grading device functionality which enables:
• Core capability: Classification of objects, detection of specific "object activity", such as "stopping", "starting", "direction of movement", etc.
Examples are listed in Annex A.
• Complex capability: Detection of "scenarios" which are based on combinations of object activity, such as "loitering", "perimeter intrusion detection", "person down", "tailgating", "intrusion", "abandoned object detection", explosion, fire, flood, potential terrorist attack using a vehicle, owner of an abandoned bag, etc.
Examples of current scenarios are listed and described in Annex B.
• Degree of difficulty: The application of real operating environments to test the performance under known or required operating stress levels, examples of operating stress levels that are sterile or non-sterile, indoor or outdoor, target obscuration levels, extreme weather conditions, vibrating mechanical rugged environments causing image shake resulting in degradation of image quality requirement, see Table 1 and Annex C.
The purpose of this document is to provide end users, at different levels of the service process, from users and installers, integrators and maintenance companies, to certification providers, with methods to measure the performance of video analysis systems that must also comply with other parts of the standard.
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This Standard specifies the functions, performance, interfaces, environmental adaptability, test methods, performance evaluation and grading rules of real-time intelligent video analysis in surveillance systems. This Standard applies to live and forensic, real-time intelligent video analysis devices and systems in video surveillance. The standard is centered on testing performance and grading device functionality which enables; - Core Capability: Classification of Objects, Detection of specific "object activity" such as "stopping", "starting", "direction of movement", etc. Examples and current expanded list in Annex H - Complex Capability: Detection of "scenarios" which are based on combinations of object activity such as "loitering", "perimeter intrusion detection", "person down", "tailgating", "intrusion", “abandoned object detection”, Explosion, Fire, Flood, Potential Terrorist attack using a vehicle, Owner of an abandoned bag, etc. Examples of current scenarios are listed and described in Annex I
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IEC 63413:2026 provides guidance and establishes requirements for the qualification of I&C platforms, according to IEC nuclear standards, aimed to be used in nuclear applications important to safety. Qualification of an I&C platform (also called pre-qualification or generic qualification) is performed as a generic activity outside the framework of a plant-specific application project. Platform qualification aims to fulfil in an anticipated manner all requirements leading to an I&C platform fully suitable for the implementation of a plant-specific application project.
The platform qualification covers consequently all I&C platform specific aspects including hardware, software, HPD, engineering tools, environmental qualification, quality and the applied safety life cycle supporting these activities. Platform qualification also considers interaction of platform assemblies. This document applies to I&C platforms under development, as well as to pre-existing I&C platforms.
This document provides guidance on what is relevant to I&C platform qualification and what is relevant to the implementation of an I&C system design based on a qualified I&C platform.
The objective of this document is to identify those requirements that relate closely to an I&C platform qualification and aims to support a two-step approach for the licensing of I&C systems based on this I&C platform. If the complete qualification of an I&C system is performed in the framework of a plant-specific application, this document does not apply.
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IEC 63423:2026 applies to cable connector assemblies for harsh environment purposes that are important to safety. It covers the engineering safety aspects to be met in the design, qualification, fabrication, assembling, testing, and installation on site of cable connector assemblies to be operated under accident conditions.
This document can be additionally used for cable assemblies that are operated under normal conditions.
This document covers cable connector assemblies that are used for the following generic functions:
- signal transmission (AC or DC voltage/current; pulses, frequency), or
- supplying electrical energy to sensors, transducers, or other devices.
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IEC 62676-6:2026 specifies the functions, performance, interfaces, environmental adaptability, test methods, performance evaluation and grading rules of real-time intelligent video analysis in surveillance systems.
This document applies to live and forensic, real-time intelligent video analysis devices and systems in video surveillance.
The document is centred on testing performance and grading device functionality which enables:
• Core capability: Classification of objects, detection of specific "object activity", such as "stopping", "starting", "direction of movement", etc.
Examples are listed in Annex A.
• Complex capability: Detection of "scenarios" which are based on combinations of object activity, such as "loitering", "perimeter intrusion detection", "person down", "tailgating", "intrusion", "abandoned object detection", explosion, fire, flood, potential terrorist attack using a vehicle, owner of an abandoned bag, etc.
Examples of current scenarios are listed and described in Annex B.
• Degree of difficulty: The application of real operating environments to test the performance under known or required operating stress levels, examples of operating stress levels that are sterile or non-sterile, indoor or outdoor, target obscuration levels, extreme weather conditions, vibrating mechanical rugged environments causing image shake resulting in degradation of image quality requirement, see Table 1 and Annex C.
The purpose of this document is to provide end users, at different levels of the service process, from users and installers, integrators and maintenance companies, to certification providers, with methods to measure the performance of video analysis systems that must also comply with other parts of the standard.
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IEC TR 63400:2025 augments that description to enable users of individual IEC SC 45A standards to obtain a more comprehensive understanding of the overall structure of the series and its relationship with other standards bodies and standards. The publication of this document and its subsequent editions should also enable minor changes in the structure to be described without the need for amending the common description that is included in the Introduction, item d), of all IEC SC 45A documents.
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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.
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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).
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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.
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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.
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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.
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This document provides information and guidelines on the decommissioning of a medical cyclotron facility, with a focus on activated or contaminated parts. Useful information and guidelines are given on decommissioning strategy and plan, safety assessment, and various decommissioning activities. This document also provides the guideline on the estimation of activation level using Monte Carlo simulation and the methodology for the measurement of activated radionuclides in the main structure, system components, and shielding walls, ceilings and floors during operation and decommissioning. Financial provisions and radioactive waste management aspects are also included. This document can be used by organizations responsible for operation and decommissioning of a medical cyclotron facility. In addition, it is expected that organizations that design a medical cyclotron or manage radioactive waste generated by cyclotron can utilize or refer to this document in whole or in part.
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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.
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IEC 63374:2025 specifies the characteristics and test methods for reactivity meters. Other methods for measuring reactivity are not addressed in this document. This document provides guidance for the design, production and operation of reactivity meters. This document is applicable to various types of nuclear reactors that can be described by the neutron kinetic point reactor model, such as pressurized water reactors (PWRs), boiling-water reactors (BWRs) or fast breeder reactors (FBRs). This document is applicable to all on-line measuring instruments that directly obtain reactivity values by measuring the neutron flux. The subject relates to the reactor nuclear parameter measurement domain.
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IEC 63435:2025 specifies the characteristics of operator support systems (OSS) used by the control room staff, maintenance engineers and emergency response staff, establishes general principles for OSS lifecycle and requirements for OSS design following the human factors engineering (HFE) programme. This document also gives the human factors guidelines and the verification and validation (V&V) requirements for OSS design.
This document is applicable to new nuclear facilities whose conceptual design is initiated after the publication of this document but it can also be used for designing OSS in existing nuclear facilities.
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