27.120 - Nuclear energy engineering
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Nuclear energy engineering
Kerntechnik
Energie nucleaire
Jedrska tehnika
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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.
- Technical report69 pagesEnglish languagesale 15% off
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.
- Standard50 pagesEnglish languagesale 10% offe-Library read for1 day
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).
- Standard27 pagesEnglish languagesale 10% offe-Library read for1 day
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 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 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 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.
- Standard37 pagesEnglish languagesale 10% offe-Library read for1 day
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.
- Standard33 pagesEnglish languagesale 15% off
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.
- Standard89 pagesEnglish languagesale 10% offe-Library read for1 day
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.
- Standard37 pagesEnglish languagesale 10% offe-Library read for1 day
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).
- Standard27 pagesEnglish languagesale 10% offe-Library read for1 day
This document specifies a method which covers the determination of Gd2O3 content in UO2 fuel pellets, by X-ray fluorescence spectrometry.
Either wave dispersion X-ray fluorescence (WD-XRF) or energy dispersion X-ray fluorescence (ED-XRF) is applicable, however, this document states a method by using WD-XRF using Gd Lα-line.
This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.
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The objective of this document is to characterize the gaseous effluents tritium and carbon-14 generated by nuclear facilities during operation and decommissioning and occurring in the same chemical species as hydrogen and carbon, e. g. as water vapour (HTO), hydrogen gas (HT, TT), carbon dioxide (14CO2), carbon monoxide (14CO), methane (CH3T, 14CH4). It concerns measurements on samples that are representative of a certain volume stream or volume of discharge during a given period of time and of the corresponding volume discharged. The result is therefore expressed in becquerels. This document applies to samples that were obtained by sampling methods according to ISO 20041-1[ REF Reference_ref_10 \r \h 9 08D0C9EA79F9BACE118C8200AA004BA90B0200000008000000110000005200650066006500720065006E00630065005F007200650066005F00310030000000 ] and describes — analysis methods for the determination of tritium and carbon-14 activities by liquid scintillation counting, and — calculation methods to determine the tritium activities discharged as tritiated water vapour (HTO) and tritium in other chemical compounds (non-HTO) as well as carbon-14 activities discharged as carbon dioxide (14CO2) and carbon-14 in other chemical compounds (non-14CO2). This document does not apply to tritium and carbon-14 activity concentrations in the environmental air, e.g. in the vicinity of nuclear installations. The accountability rules of the activities discharged necessary for the establishment of regulatory reports do not fall within the scope of this document and are the responsibility of the regulatory bodies.
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This document specifies a method which covers the determination of Gd2O3 content in UO2 fuel pellets, by X-ray fluorescence spectrometry.
Either wave dispersion X-ray fluorescence (WD-XRF) or energy dispersion X-ray fluorescence (ED-XRF) is applicable, however, this document states a method by using WD-XRF using Gd Lα-line.
This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.
- Standard15 pagesEnglish languagesale 10% offe-Library read for1 day
This standard specifies the general requirements of the corrosion control engineering life cycle in nuclear power plants. This standard applies to of various activities management of the corrosion control engineering life cycle in nuclear power plants.
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IEC 61225:2025 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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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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IEC 60911:2025 applies to pressurized water reactors (PWRs) and presents requirements for the monitoring of adequate cooling within the core in all operations, including normal and abnormal operations. Requirements for core cooling monitoring during conditions beyond a design basis accident, i.e. a design extension condition of type A or type B, are also covered in this document.
This document defines requirements for instrumentation to measure coolant parameters, which are of interest when abnormal conditions arise with either one or two phases of coolant or with gas included in the reactor pressure vessel (RPV).
This second edition cancels and replaces the first edition published in 1987. This edition includes the following significant technical changes with respect to the previous edition:
a) Modification of the title.
b) Integration and merging with the content of IEC 62117:1999 relative to the monitoring of core cooling during cold shutdown.
c) Integration of feedback following the 2011 Fukushima accident.
- Standard45 pagesEnglish languagesale 15% off
This document is applied to fuel fabrication. It describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the (U,Pu)O2 pellet microstructure.
The examinations are performed
a) before any treatment or any etching, and
b) after thermal treatment or after chemical or ion etching.
They allow
— observation of any cracks, intra- and intergranular pores or inclusions, and
— measurement of the grain size, porosity and plutonium homogeneity distribution.
The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2]. The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen.
The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching or by alpha autoradiography. A scanning electron microscope (SEM) or a microprobe can also be used. In this case an additional preparation can be needed depending on the equipment used. This preparation is not in the scope of this standard.
- Standard14 pagesEnglish languagesale 10% offe-Library read for1 day
This document is applied to fuel fabrication. It describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the (U,Pu)O2 pellet microstructure.
The examinations are performed
a) before any treatment or any etching, and
b) after thermal treatment or after chemical or ion etching.
They allow
— observation of any cracks, intra- and intergranular pores or inclusions, and
— measurement of the grain size, porosity and plutonium homogeneity distribution.
The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2]. The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen.
The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching or by alpha autoradiography. A scanning electron microscope (SEM) or a microprobe can also be used. In this case an additional preparation can be needed depending on the equipment used. This preparation is not in the scope of this standard.
- Standard14 pagesEnglish languagesale 10% offe-Library read for1 day
This document is applied to fuel fabrication. It describes the ceramographic procedure used to prepare sintered (U,Pu)O2 pellets for qualitative and quantitative examination of the (U,Pu)O2 pellet microstructure. The examinations are performed a) before any treatment or any etching, and b) after thermal treatment or after chemical or ion etching. They allow — observation of any cracks, intra- and intergranular pores or inclusions, and — measurement of the grain size, porosity and plutonium homogeneity distribution. The mean grain diameter is measured by one of the classic methods: counting (intercept method), comparison with standard grids or typical images, etc.[2]. The measurement of individual grain sizes requires uniform development of the microstructure over the entire specimen. The plutonium cluster and pore distribution and localization are generally analysed by automatic image analysis systems. The plutonium distribution is usually revealed by chemical etching or by alpha autoradiography. A scanning electron microscope (SEM) or a microprobe can also be used. In this case an additional preparation can be needed depending on the equipment used. This preparation is not in the scope of this standard.
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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.
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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 the utilization and characteristics of instrumentation used to detect seismic events at nuclear power plants with water cooled reactors. The document can also be applied to other nuclear facilities after verifying its applicability. The following types of electrical systems and equipment are not covered by this document: — seismic instrumentation involved in the implementation of nuclear safety functions as defined by IEC 61226, for example automatic shutdown systems; — seismic instrumentation not involved in the implementation of nuclear safety functions as defined by IEC 61226 but which due, for example, to close proximity to other safety classified systems, requires hardware qualification to be performed. Such systems are specified, designed, manufactured, qualified, operated and dismantled according to the relevant requirements of IEC standards, in particular IEC 61513 and the lower level IEC standards according to the safety class and technologies used. Seismic instrumentation used for the implementation of seismic reactor trip systems are developed according to the requirements of IEC 63186. An automatic shutdown system is not covered by this document. This document specifies the requirements to be fulfilled by the seismic instrumentation such that, firstly, it can be ascertained whether any of the design quantities on which the plant walk-down level and the inspection levels are based have been exceeded and that, secondly, the recording of the time history of the earthquake provides the necessary input values for a post-seismic analysis. The requirements are specified such that, independent of the detection and recording system, comparable results within tolerances are achieved in the time range as well as the frequency range.
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IEC/IEEE 62582-1:2024 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.
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- Standard21 pagesEnglish languagesale 15% off
IEC/IEEE 63332-387:2024 defines the criteria for the application and testing of diesel generator (DG) units used as safety class standby power supplies in nuclear facilities. In general, the standard applies to new nuclear facilities as well as for upgrading or back-fitting of existing facilities. Existing facilities can voluntarily adopt the requirements to enhance the performance capabilities and reliability of the installed DG units. The standard can be used in applications where highly reliable onsite alternating current (AC) power source is required to maintain plant safety following an event with potential loss of offsite power for an extended duration.
This document provides the principal design criteria, the design features, testing, and qualification requirements for the individual DG units that enable them to meet their functional requirements as a part of the standby power supply under the conditions produced by the design basis events catalogued in the plant safety analysis.
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- Amendment7 pagesEnglish languagesale 10% offe-Library read for1 day
- Amendment7 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 63351:2024 specifies the basic principles and requirements for the application of a human factors engineering (HFE) programme to the design of the human-machine interfaces (HMI) throughout the lifetime of a nuclear facility. The focus of this document is on control rooms and control functions as discussed in the text.
This document focuses on the application of a human factors engineering (HFE) programme to the design of the human-machine interfaces throughout the lifetime of a nuclear facility, including consideration of plant modifications.
This document is applicable to nuclear facilities such as: nuclear power plants (NPPs), research reactors, uranium enrichment and nuclear fuel fabrication facilities, spent fuel storage and reprocessing facilities.
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IEC TR 63486:2024 provides a cybersecurity framework for digital I&C programmable systems [2]. IEC 62645 [1] aligns strongly with the information security management system (ISMS) elements detailed within ISO/IEC 27001:2013 [2]. The ISO/IEC ISMS structure corresponds to the “I&C digital programmable system cybersecurity program” in the context (as defined in 5.2.1 of IEC 62645:2019 [1]).
The scope of this document is to capture the national and international cyber-risk approaches employed to manage cybersecurity risks associated with Instrumentation and Control (I&C) and Electrical Power Systems (EPS) at a Nuclear Power Plant (NPP).
This document summarizes an evaluation of cyber-risk approaches that are in use by nuclear facility operators to manage cybersecurity risks.
The scope of this document generally follows the exclusions of IEC 62645 which are:
- Non-malevolent actions and events such as accidental failures, human errors (except those stated above, such as impacting the performance of cybersecurity controls), and natural events. In particular, good practices for managing applications and data, including backup and restoration related to accidental failure, are out of scope.
This document summarizes key insights of the international and cyber-risk approaches used at NPPs regarding the application of ISO/IEC 27005:2018 [5]. The evaluation is based on 11 challenges to cybersecurity risk management and their applicability to NPP risk management. The challenges are detailed in Clause 7. This document also relates the risk management elements of IEC 62645 and IEC 63096.
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IEC 63272:2024 specifies the performance and functional characteristics of the on-site AC interruptible power supply systems and applies to new nuclear facilities and newly installed or upgraded on-site AC interruptible power supply systems.
The specific design requirements for the components of the power supply system are defined by the IEC standards listed in the normative references and are outside the scope of this document.
The purpose of this document is to provide high level requirements for the design of on-site AC interruptible power supply systems as part of the overall electrical distribution system in a nuclear facility. This document defines the requirements for an electrical designer to establish the design of the AC interruptible electrical power supply for nuclear facilities. It is used in conjunction with Level 1 standard IEC 63046.
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IEC 63298:2024 provide high level requirements and recommendations for the coordination of NPPs and the electric grid; see also item a) of the Introduction. The specific design requirements for components and equipment are covered by other specific IEC standards outside the scope of this document. This document also defines the coordination requirements to ensure that operating instructions for the electric grid and the NPP are developed to provide a means of safe and reliable operation. This document also defines the requirements for the development of a framework for any specific tests that may be deemed necessary for the electric grid and the NPP, such as testing of NPP regulation capabilities and load rejection to house load operation tests.
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IEC/IEEE 62582-3:2024 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 the 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. This document applies to non-energised equipment. This document is published as an IEC/IEEE Dual Logo standard. This second edition cancels and replaces the first edition published in 2012.
This edition includes the following technical changes with respect to the previous edition:
a) Updated best practices relating to condition monitoring using the tensile elongation method.
b) Updated bibliography, references and context.
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This document specifies the applicable requirements related to the design and the operation of confinement and ventilation systems for fusion facilities for tritium fuels and tritium fuel handling facilities specific for fusion applications for peaceful purposes using high tritium inventories, as well as for their specialized buildings such as hot cells, examination laboratories, emergency management centres, radioactive waste treatment and storage facilities. In most countries, a tritium quantity is declared as high for tritium inventories higher than a range of 10 g to 100 g. In the tritium fusion facilities in the scope of this document, the tritium inventory is deemed to be higher than this range for the whole site. This document applies especially to confinement and ventilation systems that ensure the safety function of nuclear facilities involved in nuclear fusion with the goal to protect the workers, the public and the environment from the dissemination of radioactive contamination originating from the operation of these installations, and in particular from airborne tritium contamination with adequate confinement systems. The types of confinement systems for other facilities are covered by ISO 26802 for fission nuclear reactors, by ISO 17873 for facilities other than fission nuclear reactors and by ISO 16647 for nuclear worksite and for nuclear installations under decommissioning. The facilities covered by these three standards, notably ISO 17873, include tritium as a radioactive material among the ones to be confined, but tritium is not their driver of the risks for workers and for members of the public. Nevertheless, the tritium quantities and risks from fusion facilities create specificities for a specific standard (e.g. in fusion facilities, tritium is the driver of routine and accident consequences). Therefore, the scope of this document does not cover the other facilities involved in tritium releases (ISO 17873, ISO 16647 and ISO 26802), even though these other facilities create tritium releases (e.g. non-reactor fission facilities, tritium laboratories, tritium removal facilities from fission plants, tritium defence facilities).
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This document specifies an analytical method for the electrochemical measurement of pure plutonium nitrate solutions of nuclear grade, with an expanded uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, k = 2). The method is applicable for aqueous solutions containing plutonium at more than 0,5 g/l and test samples containing plutonium between 4 mg and 15 mg. Application of this technique to solutions containing plutonium at less than 0,5 g/l and test samples containing plutonium at less than 4 mg requires experimental demonstration by the user that applicable data quality objectives will be met.
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This document specifies an analytical method for the electrochemical measurement of pure plutonium nitrate solutions of nuclear grade, with an expanded uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, k = 2). The method is applicable for aqueous solutions containing plutonium at more than 0,5 g/l and test samples containing plutonium between 4 mg and 15 mg. Application of this technique to solutions containing plutonium at less than 0,5 g/l and test samples containing plutonium at less than 4 mg requires experimental demonstration by the user that applicable data quality objectives will be met.
- Standard37 pagesEnglish languagesale 10% offe-Library read for1 day
This document deals with the terminological data used in the standards regarding the standardization and promotion of good practices associated with the planning, design, construction, operation and decommissioning of installations, processes and technologies involving radioactive materials. The vocabulary of nuclear installations, processes and technologies includes fuel cycle, ex-reactor nuclear criticality safety, analytical methodologies, transport of radioactive materials, materials characterization, radioactive waste management and decommissioning. NOTE See Annex A for the methodology used to develop the vocabulary.
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This document specifies an analytical method for the electrochemical measurement of pure plutonium nitrate solutions of nuclear grade, with an expanded uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, k = 2). The method is applicable for aqueous solutions containing plutonium at more than 0,5 g/l and test samples containing plutonium between 4 mg and 15 mg. Application of this technique to solutions containing plutonium at less than 0,5 g/l and test samples containing plutonium at less than 4 mg requires experimental demonstration by the user that applicable data quality objectives will be met.
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This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443.
NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
- Technical specification26 pagesEnglish languagesale 10% offe-Library read for1 day
This document specifies a method which covers the determination of Gd2O3 content in UO2 fuel pellets, by X-ray fluorescence spectrometry. Either wave dispersion X-ray fluorescence (WD-XRF) or energy dispersion X-ray fluorescence (ED-XRF) is applicable, however, this document states a method by using WD-XRF using Gd Lα-line. This method has been tested for mass fractions of from 2 % to 10 % Gd2O3.
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This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443.
NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
- Technical specification26 pagesEnglish languagesale 10% offe-Library read for1 day
This document complements the existing requirements of ISO/IEC 17021-1 for bodies providing audit and certification of quality management systems against ISO 19443. NOTE This document can be used as a criteria document for accreditation, peer assessment or other audit processes.
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This document applies to nuclear power plants with water cooled reactors. For other nuclear facilities check the applicability of the document in advance, before it might be applied correspondingly. This document specifies the requirements for the earthquake safety of components. The operation-specific safety-related requirements for each component, e.g. load-bearing capacity (stability), integrity and functionality (see 4.1) are not the subject of this document. With regard to analysing the mechanical behaviour of the individual components and verifying the fulfillment of their safety related functions, additionally, the respective component-specific standards need to be consulted. In this document, the term "mechanical components" refers to components such as vessels, heat exchangers, pumps, valves, lifting gear, distribution systems and pipe lines including their support structures in as far as these components are not considered to be civil structures in accordance with ISO 4917-3. Liners, crane runways, platforms and scaffoldings are not considered as being part of these mechanical components. In this document, the term electrical components refers to the combination of electrical devices including all electrical connections and their support structures (e.g. cabinets, frames, consoles, brackets, suspensions or supports). Supplementary to this document the seismic qualification of electrical components is reported in IEC/IEEE 60980-344. NOTE This document is independent of national standards. Recommendations, given in Annex A, are mainly based on the Eurocodes-Design-Philosophy and European Standards. Alternatively other equivalent standards or regulations can be used in case the general requirements given in this document together with Annex A can be met.
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This document applies to nuclear power plants with water cooled reactors. This document does not apply to earthquakes stronger than the design basis earthquake. This document specifies guidance on the actions to be taken in preparation for and following an earthquake at a nuclear power plant. This document is intended to be used as a guideline for decision making regarding continued operation, shutdown and restart of the nuclear power plant after an earthquake. It can also be used to assist operating organizations in the preparation and implementation of an overall pre- and post-earthquake action programme for dealing with situations in accordance with the level of seismic ground motion experienced at the site, and the seismic design level of the plant.
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- Standard1 pageFrench languagesale 15% off
This document applies to nuclear power plants with water cooled reactors and, in particular, to the design of components and civil structures against seismic events in order to meet the safety objectives. For other nuclear facilities the applicability of the document is checked in advance, before it might be applied correspondingly. Seismic isolation is not adressed in the series of ISO 4917. The following safety objectives are defined in order to ensure the protection of people and the environment against radiation risks: a) controlling reactivity; b) cooling fuel assemblies; c) confining radioactive substances; d) limiting radiation exposure.
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This document applies to civil structures of nuclear power plants with water cooled reactors in order to achieve the safety objectives given in ISO 4917-1. For other nuclear facilities the applicability of the document needs to be checked in advance, before it might be applied correspondingly. This document specifies the requirements for civil structures for the verification of their load-bearing capacity in case of a seismic event. Additionally, requirements are specified pertaining to the verification of the serviceability of civil structures as far as necessary for maintaining their safety-related function in case of a seismic event (e.g. deformation and crack-width limitations). This document will be applied under the presumption that the geology and tectonics of the plant site have been investigated with special emphasis on the existence of active geological faults and lasting geological ground displacements, and that the site has been deemed suitable for a nuclear installation. To achieve these goals, this document deals with the requirements specific to the seismic design of civil structures above and beyond their conventional design. The basic requirements of these precautionary measures are dealt with in ISO 4917-1. This document does not apply to cranes, to detachment devices for lifting equipment nor to the supporting and mounting constructions of components. This document is independent of national standards. Recommendations, given in Annex A, are mainly based on the KTA Design-Philosophy and European standards. Alternatively other equivalent standards or regulations can be used in case the general requirements given in this document can be met. NOTE The term civil structures as used in this document comprise buildings and structural members made of reinforced concrete, pre-stressed concrete, steel, as well as steel composite structures and masonry. Among others, these include the containment, crane runways, platforms, fastening constructions and canals.
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This document describes methodologies for radioactivity characterization of very low-level waste (VLLW) generated from the operation or decommissioning of nuclear facilities. The purpose is to differentiate VLLW from low-level radioactive solid waste and waste below clearance levels. The aim is to effectively characterize and to demonstrate that it satisfies the criteria for VLLW. This document focuses specifically on characterization methods of radioactive solid waste. Clearance and exemption monitoring are not covered within this document. Additionally, the characterization of liquid and gaseous wastes is also excluded from this document.
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