27.120 - Nuclear energy engineering
ICS 27.120 Details
Nuclear energy engineering
Kerntechnik
Energie nucleaire
Jedrska tehnika
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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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ISO 12183:2016 describes an analytical method for the electrochemical assay of pure plutonium nitrate solutions of nuclear grade, with a total uncertainty not exceeding ±0,2 % at the confidence level of 0,95 for a single determination (coverage factor, K = 2). The method is suitable for aqueous solutions containing more than 0,5 g/L plutonium and test samples containing between 4 mg and 15 mg of plutonium. Application of this technique to solutions containing less than 0,5 g/L and test samples containing less than 4 mg of plutonium requires experimental demonstration by the user that applicable data quality objectives will be met.
For some applications, purification of test samples by anion exchange is required before measurement to remove interfering substances when present in significant amounts.
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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 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 is recommended for use 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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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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This document is applicable to the determination of gadolinium as Gd2O3 in powder blends and sintered
pellets of Gd2O3 + UO2 and ((U, Gd) O2) from mass fraction 10 g/kg to 100 g/kg (i.e. 1 % to 10 %), using
a suitable ICP-AES instrument.
This methodology is capable of demonstrating compliance with agreed upon fuel specifications and
associated data quality objectives provided the user has performed qualification measurements
under their established measurement control program to demonstrate that measurement uncertainty
requirements will be met with the desired level of confidence at the specification
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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 TR 63415:2023 provides an overview over the formalized modelling and designing of cybersecure architectures to apply for I&C system cybersecurity enforcement at NPPs. The plant-specific risk assessment can use the techniques covered by this TR. This document considers the complex problem of NPP I&C architecture synthesis to address particular issues:
- asset classification,
- barrier measures assignment,
- the information transfer and links conformity with security requirements.
This document provides guidance on creating a comprehensive security model applicable to NPP I&C systems that describes NPP I&C cybersecurity architecture and aids in accomplishing the main tasks of I&C system secure design, which are:
- specification of system designs with increased determinism that enhance security,
- mapping of the security requirements into the security architecture of the I&C system,
- definition of the security requirements for information exchange between components within the I&C system, operators and other systems,
- assistance in the determination of the security degree assignment with a model-based technique considering asset properties and formal grouping of the assets,
design and establishment of security zones boundaries.
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This document specifies a method for the determination of uranium concentrations in nitric acid or TBP-DILUANT (for example TBP-kerosene) solutions coming from the nuclear fuel cycle.
The method is applicable
— for process control of solutions, free of suspension, which contain between 10 g/l to 300 g/l uranium,
and
— for high accuracy purposes (Safeguards) to nitric acid solutions, free of suspension, which contain
between 100 g/l and 220 g/l uranium.
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This document is applicable to the determination of gadolinium as Gd2O3 in powder blends and sintered
pellets of Gd2O3 + UO2 and ((U, Gd) O2) from mass fraction 10 g/kg to 100 g/kg (i.e. 1 % to 10 %), using
a suitable ICP-AES instrument.
This methodology is capable of demonstrating compliance with agreed upon fuel specifications and
associated data quality objectives provided the user has performed qualification measurements
under their established measurement control program to demonstrate that measurement uncertainty
requirements will be met with the desired level of confidence at the specification
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This document specifies a method for the determination of uranium concentrations in nitric acid or TBP-DILUANT (for example TBP-kerosene) solutions coming from the nuclear fuel cycle.
The method is applicable
— for process control of solutions, free of suspension, which contain between 10 g/l to 300 g/l uranium,
and
— for high accuracy purposes (Safeguards) to nitric acid solutions, free of suspension, which contain
between 100 g/l and 220 g/l uranium.
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IEC TR 63468:2023 overviews the fundamentals of artificial intelligence (AI) as it could potentially be applied within nuclear facilities and identifies proven or potential applications, with the objective to foster better understanding and adoption of AI technologies within such facilities. With the objective of supporting future standard development work of IEC SC 45A in this technical area, this document takes the initiative to propose a structure for SC 45A standard series on nuclear AI applications and recommends setting up a new dedicated working group to be responsible for and coordinate standard development efforts in this particular area, taking into account its cross-cutting nature.
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This document specifies the methods and techniques for leak tightness assessment of a metallic component at high temperature by measuring its total leakage rates in a vacuum chamber with a tracer gas leak detector and high-pressure helium gas or the gas mixture flowing out of the component as tracer gas during its thermal and pressure cycles at its operating conditions. The minimum detectable leakage rate can be as low as 10-10 Pa·m3/s, depending on the dimension, external configuration complexity and materials of the component, and is strongly related to the test system and the test conditions. This document is applicable for the hot helium leak test of in-vessel components as per its normal operating conditions in nuclear fusion reactors, which operate at elevated temperatures in an ultra-high vacuum environment down to 10-6 Pa and with inner flowing-coolant at operating pressure. It is also applicable to the overall leak tightness test of welds in other metallic components and equipment that could be evacuated and pressurized, such as pressurized tanks, pipes and valves in power plants, aerospace and other nuclear reactors.
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This document is the first of a series of seven documents which outlines the general principles to manage the various type of radioactive waste, and provides guidance for the practical implementation of those principles. The purpose of this document is to address the following: a) principles, objectives and practical approaches for radioactive waste management; b) outline of the structure of series from ISO 24389-1 through ISO 24389-7.
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This document specifies minimum nuclear criticality safety training requirements for operations staff, operations supervisors, and management.
This document is applicable to areas, processes or facilities containing quantities of fissile material for which nuclear criticality safety assessment is required as defined in ISO 1709.
This document is not applicable to the transport of fissile materials outside the boundaries of nuclear establishments.
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This document is applicable to housed scintillators for registration and spectrometry of alpha-, beta-, gamma-, X-ray and neutron radiation. Their basic parameters such as a light output and intrinsic resolution are established. The document does not apply to gas or liquid scintillators and scintillators for counting or
current measurement.
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- Amendment6 pagesEnglish languagesale 10% offe-Library read for1 day
This document specifies minimum nuclear criticality safety training requirements for operations staff, operations supervisors, and management.
This document is applicable to areas, processes or facilities containing quantities of fissile material for which nuclear criticality safety assessment is required as defined in ISO 1709.
This document is not applicable to the transport of fissile materials outside the boundaries of nuclear establishments.
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The scope of ISO 16659 series is to provide different test methods aiming at assessing the efficiency of radioactive iodine traps in ventilation systems of nuclear facilities. The ISO 16659 series deals with iodine traps containing a solid sorbent — mainly activated and impregnated charcoal, the most common solid iodine sorbents used in the ventilation systems of nuclear facilities — as well as other sorbents for special conditions (e.g. high temperature zeolites). The scope of this document is to provide general and common requirements for the different test methods for industrial nuclear facilities. The different methods will be described in other specific parts of ISO 16659 series. Nuclear medicine applications are excluded from the scope of ISO 16659 series. In principle, ISO 16659 series is used mainly for filtering radioactive iodine, but other radioactive gases can also be trapped together with iodine. In such a case, some specificity may have to be adapted for these other radioactive gases in specific parts of ISO 16659 series. This document describes the main general requirements in order to check in situ the efficiency of the iodine traps, according to test conditions that are proposed to be as reproducible as possible.
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This document is applicable to housed scintillators for registration and spectrometry of alpha-, beta-, gamma-, X-ray and neutron radiation. Their basic parameters such as a light output and intrinsic resolution are established. The document does not apply to gas or liquid scintillators and scintillators for counting or current measurement.
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- Amendment6 pagesEnglish languagesale 10% offe-Library read for1 day
This document applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR. This document specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example, the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed).
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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, 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. This method does not generate a toxic mixed waste as does the potassium dichromate titration in ISO 7097-1.
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IEC 60951-1:2022 is available as IEC 60951-1:2022 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.IEC 60951-1:2022 provides general guidance on the design principles and performance criteria for equipment to measure radiation and fluid (gaseous effluents or liquids) radioactivity levels at nuclear facilities during and after design basis accidents (DBA) and design extension conditions (DEC), including severe accident (SA). This document is limited to equipment for continuous monitoring of radioactivity in design basis accidents (DBA), design extension conditions (DEC), including severe accident (SA) and post-accident conditions. The purpose of this document is to lay down general requirements and give examples of acceptable methods for equipment for continuous monitoring of radioactivity within the facility during and after design basis accidents (DBA), design extension conditions (DEC), including severe accident (SA) in nuclear facilities. This third edition cancels and replaces the second edition published in 2009.The main technical changes with regard to the previous edition are as follows.
- title modified.
- to be consistent with the categorization of the accident condition.
- to update the references to new standards published since the second edition.
- to update the terms and definitions.
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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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IEC 62705:2022 is available as IEC 62705:2022 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.
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 document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements.
It does not add to, subtract from, or in any way modify those requirements.
This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.
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IEC 60910:2022 provides requirements for primary and secondary containment parameter monitoring that enable the operator to identify developing deviations from normal operation. The operator can then take corrective action at an early stage to prevent a minor failure from developing into a serious plant failure or an accident condition. This document is directed towards monitoring the primary and secondary containment under normal conditions only.
This second edition cancels and replaces the first edition, published in 1988. This edition includes the following significant technical changes with respect to the previous edition:
a. Modification of title;
b. Integration of new technology and knowledge;
c. Drafting directed towards monitoring conditions in containment under normal conditions.
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IEC 60951-3:2022 is available as IEC 60951-3:2022 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.
IEC 60951-3:2022 provides general guidance on the design principles and performance criteria for equipment for continuous high range area gamma monitoring in nuclear facilities for accident and post-accident conditions. This document categorizes accident conditions into design basis accidents (DBA) and design extension conditions (DEC), including severe accident (SA). The purpose of this document is to lay down general requirements for equipment for continuous high range area gamma monitoring of radiation within the facility during and after accident conditions in nuclear facilities. This document is applicable to installed dose rate meters that are used to monitor high levels of gamma radiation during and after an accident. This third edition cancels and replaces the second edition published in 2009.The main technical changes with regard to the previous edition are as follows:
- Title modified.
- To be consistent with the categorization of the accident condition.
- To update the references to new standards published since the second edition.
- To update the terms and definitions.
This standard is to be read in conjunction with IEC 60951-1.
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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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IEC/IEEE 62582-4:2022 is available as IEC/IEEE 62582-4:2022 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.IEC/IEEE 62582-4:2022 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. This second edition cancels and replaces the first edition published in 2011, and its Amendment 1:2016. This edition includes the following significant technical changes with respect to the previous edition:
- Consideration of publication of IEC/IEEE 60780-323;
- An example added in Annex B and update;
- Annex C added.
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IEC/IEEE 62582-2:2022 is available as IEC/IEEE 62582-2:2022 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.
IEC/IEEE 62582-2:2022 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. This document is intended for application 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 2011, and its Amendment 1:2016. This edition includes the following significant technical changes with respect to the previous edition:
- Modification of the title;
- Consideration of publication of IEC/IEEE 60780-323.
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This document provides guidance on the implementation of the ISO 19443 requirements, with examples of possible steps an organization can take to meet the requirements.
It does not add to, subtract from, or in any way modify those requirements.
This document does not prescribe mandatory approaches to implementation, or provide any preferred method of interpretation.
- Technical report65 pagesEnglish languagesale 10% offe-Library read for1 day
This document specifies requirements for the software of computer-based instrumentation and control (I&C) systems performing functions of safety category B or C as defined by IEC 61226. It complements IEC 60880 which provides requirements for the software of computer-based I&C systems performing functions of safety category A. It is consistent with, and complementary to, IEC 61513. Activities that are mainly system level activities (for example, integration, validation and installation) are not addressed exhaustively by this document: requirements that are not specific to software are deferred to IEC 61513. The link between functions categories and system classes is given in IEC 61513. Since a given safety-classified I&C system may perform functions of different safety categories and even non safety-classified functions, the requirements of this document are attached to the safety class of the I&C system (class 2 or class 3). This document is not intended to be used as a general-purpose software engineering guide. It applies to the software of I&C systems of safety classes 2 or 3 for new nuclear power plants as well as to I&C upgrading or back-fitting of existing plants. For existing plants, only a subset of requirements is applicable and this subset has to be identified at the beginning of any project. The purpose of the guidance provided by this document is to reduce, as far as possible, the potential for latent software faults to cause system failures, either due to single software failures or multiple software failures (i.e. Common Cause Failures due to software). This document does not explicitly address how to protect software against those threats arising from malicious attacks, i.e. cybersecurity, for computer-based systems. IEC 62645 provides requirements for security programmes for computer-based systems.
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This document establishes requirements relevant to the selection and use of wireless devices
in instrumentation and control (I&C) systems important to safety used in nuclear power plants
(NPPs). Those I&C systems may fully consist of wireless devices.
NOTE The word “use” refers to the integration of the device, its qualification, administrative control, and every
other activity that may be necessary to use the device in an important to safety application.
This document applies to the I&C of new NPPs and to backfit of I&C in existing NPPs. Every
wireless device or wireless system that is important to safety is in the scope of this document.
Both fixed and mobile devices and all data types (voice, process data, etc.) are included
within the scope if they provide a safety classified function.
This document restricts the use of wireless devices to systems supporting category C
functions according to IEC 61226, excluding explicitly their use for categories A and B.
Non-safety devices and systems may use this document as guidelines, for example to ensure
that important to safety devices are not disturbed.
– Clause 5 describes the fundamental requirements regarding safety and cybersecurity.
– Clause 6 gives wireless-specific requirements that have to be included in the system
design.
– Clause 7 describes the requirements for the selection and integration of wireless devices.
– Clause 8 deals with electromagnetic compatibility and spectrum management.
– Clause 9 gives wireless-specific requirements regarding cybersecurity.
– Clause 10 describes the requirements for the qualification of wireless devices and their
environment.
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This document applies to the design, location and application of installed equipment for
monitoring local gamma radiation dose rates within nuclear facilities during normal operation
and anticipated operational occurrences. High range area gamma radiation dose rate
monitoring equipment for accident conditions currently addressed by IEC 60951-1 and
IEC 60951-3 is not within the scope of this document.
This document does not apply to the measurement of neutron dose rate. Additional equipment
for neutron monitoring may be required, depending on the plant design, if the neutron dose
rate makes a substantial contribution to the total dose equivalent to personnel.
This document provides guidelines for the design principles, the location, the application,
the calibration, the operation, and the testing of installed equipment for continuously
monitoring local gamma radiation dose rates in nuclear facilities under normal operation
conditions and anticipated operational occurrences. These instruments are normally referred
to as area radiation monitors. Portable instruments are also used for this purpose but are not
covered by this document.
Radiation monitors utilized in area radiation monitoring equipment are addressed in
IEC 60532. As discussed in IEC 60532, measurement of gamma radiation may be expressed
by a number of alternative quantities depending on national regulations. However, for this
type of instrument, the most likely quantity to be measured is the air kerma (Gy), or the
ambient dose equivalent H*(10)(Sv).
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This document specifies:
a) the determination of mass gain;
b) the surface inspection of products of zirconium and its alloys when corrosion is tested in water at 360 °C or in steam at or above 400 °C;
c) the performance of tests in steam at 10,3 MPa.
This document is applicable to wrought products, castings, powder metallurgy products and weld metals.
This method has been widely used in the development of new alloys, heat-treating practices and for the evaluation of welding techniques. It is applicable for use in its entirety to the extent specified for a product acceptance test, rather than merely a means of assessing performance in service.
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This International Standard specifies requirements for a quality management system when an organization:
a) needs to demonstrate its ability to consistently provide products and services that meet customer and applicable statutory and regulatory requirements, and
b) aims to enhance customer satisfaction through the effective application of the system, including processes for improvement of the system and the assurance of conformity to customer and applicable statutory and regulatory requirements.
All the requirements of this International Standard are generic and are intended to be applicable to any organization, regardless of its type or size, or the products and services it provides.
NOTE 1 In this International Standard, the terms "product" or "service" only apply to products and services intended for, or required by, a customer.
NOTE 2 Statutory and regulatory requirements can be expressed as legal requirements.
This International Standard applies to organizations supplying ITNS products or services.
Application of this standard to organizations performing activities on a licensed nuclear site is subject to prior agreement by the Licensee.
Requirements specified in this International Standard are complementary (not alternative) to customer and applicable statutory and regulatory requirements.
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