27.120.20 - Nuclear power plants. Safety
ICS 27.120.20 Details
Nuclear power plants. Safety
Kernkraftwerke. Sicherheitsaspekte der Kerntechnik
Centrales nucléaires. Sécurité
Jedrske elektrarne. Varnost
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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 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 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 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 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 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.
- Technical report56 pagesEnglish languagesale 15% off
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 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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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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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 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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See the scope of IEC 62988:2018. Adoption of IEC 62988:2018 is to be done without modification.
- Standard21 pagesEnglish languagesale 10% offe-Library read for1 day
This document presents the methods and provisions for sampling tritium and carbon‑14 in the gaseous effluents generated by nuclear facilities during operation and decommissioning. Specifically included are sample withdrawal location, extraction, transport flow measurement, and collection for later analysis. This document doesn’t address to real time measurements of tritium activity and carbon-14 activity in the effluent air of stacks and ducts. Information about real time measurements can be found in ISO 2889:2021, Annex H. Sample processing, analysis and calculations of tritium and carbon‑14 emissions will be addressed in future parts of ISO 20041.
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See the scope of IEC 62465:2010. Adoption of IEC 62465 is to be done without modification.
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See the scope of the revised IEC 60709 in 45A/1113/CDV that was unchanged for the preparation of the proposal of FDIS to be circulated in parallel in CENELEC.
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See the technical scope of the amendment of IEC 62808 in 45A/1134/CDV that was unchanged for the preparation of the proposal of FDIS to be circulated in parallel in CENELEC.
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See the scope of IEC 62646:2016. Adoption of IEC 62646 is to be done without modification
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IEC 60964:2018 is available as IEC 60964:2018 RLV which contains the International Standard and its Redline version, showing all changes of the technical content compared to the previous edition.IEC 60964:2018 establishes requirements for the human-machine interface in the main control rooms of nuclear power plants. The document also establishes requirements for the selection of functions, design consideration and organization of the human-machine interface and procedures which are used systematically to verify and validate the functional design. These requirements reflect the application of human factors engineering principles as they apply to the human-machine interface during plant operational states and accident conditions (including design basis and design extension conditions), as defined in IAEA SSR-2/1 and IAEA NP-T-3.16. This third edition cancels and replaces the second edition published in 2009. This edition constitutes a technical revision. This edition includes the following significant technical changes with respect to the previous edition: a) to review the usage of the term “task” ensuring consistency between IEC 60964 and IEC 61839; b) to clarify the role, functional capability, robustness and integrity of supporting services for the MCR to promote its continued use at the time of a severe accident or extreme external hazard; c) to review the relevance of the standard to the IAEA safety guides and IEC SC 45A standards that have been published since IEC 60964:2009 was developed; d) to clarify the role and meaning of “task analysis”, e) to further delineate the relationships with derivative standards (i.e. IEC 61227, IEC 61771, IEC 61772, IEC 61839, IEC 62241 and others of relevance to the control room design); f) to consider its alignment with the Human Factors Engineering principles, specifically with the ones of IAEA safety guide on Human Factors (DS-492) to be issued.
- Standard46 pagesEnglish languagesale 10% offe-Library read for1 day
See the scope of the revised IEC 60709 in 45A/1113/CDV that was unchanged for the preparation of the proposal of FDIS to be circulated in parallel in CENELEC.
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This document specifies the requirements applicable to the design and use of airborne confinement systems that ensure safety and radioprotection functions in nuclear worksites and in nuclear installations under decommissioning to protect from radioactive contamination produced: aerosol or gas.
The purpose of confinement systems is to protect the workers, members of the public and environment against the spread of radioactive contamination resulting from operations in nuclear worksites and from nuclear installations under decommissioning.
The confinement of nuclear worksites and of nuclear installations under decommissioning is characterized by the temporary and evolving (dynamic) nature of the operations to be performed. These operations often take place in area not specifically designed for this purpose.
This document applies to maintenance or upgrades at worksites which fit the above definition.
NOTE The requirements for the design and use of ventilation and confinement systems and for liquid confinement in nuclear reactors or in nuclear installations other than nuclear worksites and nuclear installations under decommissioning are developed in other ISO standards.
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IEC TR 63400:2021 is intended to augment 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. This document outlines the scope of the IEC SC 45A standards series and describes the basic structure of the IEC SC 45A standards series, with particular reference to a hierarchy of levels and subdivision into a set of broad topic areas. it presents the structure of the IEC SC 45A standards series in diagrammatic form.
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IEC 60987:2021 provides requirements and recommendations for the hardware aspects of I&C systems whatever the technology and applies for all safety classes in a graded manner (as defined by IEC 61513). The requirements defined within this document guide, in particular, the selection of pre-existing components, hardware aspects of system detailed design and implementation and equipment manufacturing.
This third edition cancels and replaces the second edition published in 2007. This edition includes the following significant technical changes with respect to the previous edition:
a) Title modified;
b) Take account of the fact that hardware requirements apply to all I&C technologies, including conventional hardwired equipment, programmable digital equipment or by using a combination of both types of equipment;
c) Align the standard with the new revisions of IAEA documents SSR-2/1, which include as far as possible an adaptation of the definitions;
d) Replace, as far as possible, the requirements associated with standards published since the edition 2.1, especially IEC 61513, IEC 60880, IEC 62138, IEC 62566 and IEC 62566‑2;
e) Review the existing requirements and update the terminology and definitions;
f) Extend the scope of the standard to all hardware (computerized and non-computerized) and to all safety classes 1, 2 and 3;
g) Complete, update the IEC and IAEA references and vocabulary;
h) Check possible impact of other IAEA requirements and recommendations considering extension of the scope of SC 45A;
i) Highlight the use of IEC 62566 and IEC 62566-2 for HPD development;
j) Introduce specific activities for pre-existing items (selection, acceptability and/or mitigation);
k) Introduce clearer requirements for electronic module-level design, manufacturing and control;
l) Complete reliability assessment methods;
m) Introduce requirements when using automated tests or control activities;
n) Complete description of manufacturing control activities (control process, assessment of manufactured equipment, preservation of products);
o) Define and ensure the inclusion of a graded approach for dealing with the 3 different classes of equipment and related requirements.
- Standard53 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 60987:2021 provides requirements and recommendations for the hardware aspects of I&C systems whatever the technology and applies for all safety classes in a graded manner (as defined by IEC 61513). The requirements defined within this document guide, in particular, the selection of pre-existing components, hardware aspects of system detailed design and implementation and equipment manufacturing. This third edition cancels and replaces the second edition published in 2007. This edition includes the following significant technical changes with respect to the previous edition: a) Title modified; b) Take account of the fact that hardware requirements apply to all I&C technologies, including conventional hardwired equipment, programmable digital equipment or by using a combination of both types of equipment; c) Align the standard with the new revisions of IAEA documents SSR-2/1, which include as far as possible an adaptation of the definitions; d) Replace, as far as possible, the requirements associated with standards published since the edition 2.1, especially IEC 61513, IEC 60880, IEC 62138, IEC 62566 and IEC 62566‑2; e) Review the existing requirements and update the terminology and definitions; f) Extend the scope of the standard to all hardware (computerized and non-computerized) and to all safety classes 1, 2 and 3; g) Complete, update the IEC and IAEA references and vocabulary; h) Check possible impact of other IAEA requirements and recommendations considering extension of the scope of SC 45A; i) Highlight the use of IEC 62566 and IEC 62566-2 for HPD development; j) Introduce specific activities for pre-existing items (selection, acceptability and/or mitigation); k) Introduce clearer requirements for electronic module-level design, manufacturing and control; l) Complete reliability assessment methods; m) Introduce requirements when using automated tests or control activities; n) Complete description of manufacturing control activities (control process, assessment of manufactured equipment, preservation of products); o) Define and ensure the inclusion of a graded approach for dealing with the 3 different classes of equipment and related requirements.
- Standard53 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 63186:2021 specifies the minimum requirements for the design of the seismic trip system, and the components thereof, used in a nuclear power plant to mitigate seismic effects. This system is intended to shut down the reactor in operation automatically before it is significantly impacted by the vibratory ground motion incurred by strong earthquakes. This document is applicable to both the design of new built plants and the upgrading of plants in operation. It may be used for the design of other types of nuclear facilities where normal operation shall be stopped in case of strong seismic motions.
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This document specifies the requirements applicable to the design and use of airborne confinement systems that ensure safety and radioprotection functions in nuclear worksites and in nuclear installations under decommissioning to protect from radioactive contamination produced: aerosol or gas.
The purpose of confinement systems is to protect the workers, members of the public and environment against the spread of radioactive contamination resulting from operations in nuclear worksites and from nuclear installations under decommissioning.
The confinement of nuclear worksites and of nuclear installations under decommissioning is characterized by the temporary and evolving (dynamic) nature of the operations to be performed. These operations often take place in area not specifically designed for this purpose.
This document applies to maintenance or upgrades at worksites which fit the above definition.
- Standard43 pagesEnglish languagesale 10% offe-Library read for1 day
This document presents the requirements for the on-site emergency response facilities
(referred to hereinafter as the “ERF”) which are to be used in case of incidents or accidents
occurring on the associated Nuclear Power Plant (NPP). The ERF consists of the Emergency
Response Centre (ERC), the Technical Support Centre (TSC) and the Operational Support
Centre (OSC), as shown in Figure 1.
It establishes requirements for the ERF features and ERF I&C equipment to:
• coordinate on-site operational efforts with respect to safety and radioprotection;
• optimize the design in terms of environment control, lighting, power supplies and access
control of the ERF;
• enhance the identification and resolution of potential conflicts between the traditional
operational means and emergency means (MCR/SCR and ERF, operating staff and
emergency teams, operational procedures and emergency procedures);
• aid the identification and the enhancement of the potential synergies between the
traditional operational means and emergency means.
This document is intended for application to new nuclear power plants whose conceptual
design is initiated after the publication of this document, but it may also be used for designing
and implementing ERF in existing nuclear power plants or in any other nuclear facility.
Detailed equipment design is outside the scope of this document.
This document does not define the situations (reactor plant conditions, hazards and
magnitudes of hazards) leading to mobilisation of emergency response teams and activation /
use of the ERF. These aspects are usually addressed in the NPP Emergency Plan. However,
the need for consistency of the ERF design and operation with the NPP Emergency Plan is
within scope.
- Standard29 pagesEnglish languagesale 10% offe-Library read for1 day
This part of IEC/IEEE 62582 contains methods for condition monitoring of organic and
polymeric materials in instrumentation and control cables using insulation resistance
measurements in the detail necessary to produce accurate and reproducible results during
simulated accident conditions. It includes the requirements for the measurement system and
measurement procedure, and the reporting of the measurement results.
NOTE Measurement of insulation resistance during simulated accident conditions with the aim of determining the
lowest value during the accident in order to assess cable performance involves special requirements given in this
document. Methods for measurement under stable (non-accident) conditions are available in other international
standards, e.g. IEC 62631-3-3.
The different parts of the IEC/IEEE 62582 series 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 qualification of equipment important to safety is found in IEC/IEEE 60780-323.
- Standard27 pagesEnglish languagesale 10% offe-Library read for1 day
This document establishes, for nuclear power plants2, a method of assignment of the
functions specified for the plant into categories according to their importance to safety.
Subsequent classification of the I&C and electrical power systems performing or supporting
these functions, based on the assigned category, then determines relevant design criteria.
The design criteria, when applied, ensure the achievement of each function in accordance to
its importance to safety. In this document, the criteria are those of functionality, reliability,
performance, environmental qualification (e.g. seismic) and quality assurance (QA).
This document is applicable to:
- the functions important to safety that are performed by I&C systems and supported by
electrical power systems (categorization of I&C functions),
- the I&C systems that enable those functions to be implemented (classification of I&C
systems),
- the electrical power systems that support those functions (classification of electrical power
systems).
The systems under consideration provide automated protection, closed or open loop control,
information to the operating staff, and electrical power supply to systems. These systems
keep the NPP conditions inside the safe operating envelope and provide automatic actions, or
enable manual actions, that prevent or mitigate accidents, or that prevent or minimize
radioactive releases to the site or wider environment. The I&C and electrical power systems
that fulfil these roles safeguard the health and safety of the NPP operators and the public.
This document follows the general principles given in IAEA Safety Requirement SSR-2/1 and
Safety Guides SSG-30, SSG-34 and SSG-39, and it defines a structured method of applying
the guidance contained in those codes and standards to the I&C and electrical power systems
that perform functions important to safety in a NPP. This document is read in association with
the IAEA guides together with IEC 61513 and IEC 63046 in implementing the requirements of
the IEC 61508 series. The overall classification scheme of structures, systems and
components for NPPs can be summarized as follows by Figure 1.
- Standard43 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 62855 provides the electrotechnical engineering guidelines for analysis of AC and DC
electrical power systems in nuclear power plants (NPPs) in order to demonstrate that the
power sources and the distribution systems have the capability for safe operation and shut
down of the NPP, bringing it to a controlled state after an anticipated operational occurrence
or accident conditions and finally reaching a safe state.
The analytical studies discussed in this document provide assurance that the design bases
are satisfied to meet their functional requirements under the conditions produced by the
applicable design basis events. The studies provide assurance that the electrical power
system is capable of supporting safety functions during all required plant conditions.
NOTE The safety functions are described in IAEA Specific Safety Requirements SSR-2/1 related to the design of
the nuclear power plants..
Analytical studies validate the robustness and adequacy of design margins and demonstrate
the capability of electrical power systems to support plant operation for normal, abnormal,
degraded and accident conditions.
The analyses are used to verify that the electrical power system can withstand minor
disturbances and that the consequences of major disturbances or failures do not degrade the
capability of the electrical power systems to support safe shutdown of the plant and maintain
the plant in shutdown condition.
The analyses are performed with one or more of
• simulation tools (software and hardware) that have been verified and validated,
• hand calculations, and
• tests.
This document provides guidance on the types of analyses required to demonstrate that the
plant's auxiliary power system can perform the required safety functions. This document does
not provide specific details on how the analysis should be conducted.
This document does not cover digital controllers (such as controllers for rectifiers, inverters,
sequencers and electrical protection devices) used in electrical power systems. IEC 61513
gives recommendations that apply to the electronic controls and protective elements of the
electrical power systems.
This document does not include environmental conditions (i.e. temperature, humidity, etc.) or
external events (seismic, flooding, fire, high energy electromagnetic pulse, etc.) that may
impact equipment sizing or protection requirements. The external events lightning and
geomagnetic storms are included.
This document does not cover additional or unique requirements for stand-alone power
system, such as power supplies for security measures in NPPs. Pertinent clauses of this
document may be used as a guideline for such systems.
Redundancy in the power system design can increase the availability of electrical power to
critical plant equipment. Performing a probabilistic risk assessment (PRA) is a method of
assessing system availability and optimizing design for high reliability. This document does
not cover improving the reliability of NPP electrical power systems using statistical or diverse
and redundant schemes.
Requirements for safeguards of personnel involved with installation, maintenance and
operation of electrical systems and general personal safety are outside the scope of this
document. General guidance for lightning protection of equipment is provided in relevant
clauses of this document.
This document is intended to be used:
• for verification of the design of new nuclear power plants,
• for demonstrating the adequacy and impact of major modifications of electrical power
systems in operating nuclear power plants, and
• where there is a requirement to assess and establish operating limits and constraints for
existing plants.
Pertinent parts of this document can be used as guidance for decommissioning stages.
- Standard60 pagesEnglish languagesale 10% offe-Library read for1 day
This document presents the requirements for the on-site emergency response facilities (referred to hereinafter as the “ERF”) which are to be used in case of incidents or accidents occurring on the associated Nuclear Power Plant (NPP). The ERF consists of the Emergency Response Centre (ERC), the Technical Support Centre (TSC) and the Operational Support Centre (OSC), as shown in Figure 1. It establishes requirements for the ERF features and ERF I&C equipment to: • coordinate on-site operational efforts with respect to safety and radioprotection; • optimize the design in terms of environment control, lighting, power supplies and access control of the ERF; • enhance the identification and resolution of potential conflicts between the traditional operational means and emergency means (MCR/SCR and ERF, operating staff and emergency teams, operational procedures and emergency procedures); • aid the identification and the enhancement of the potential synergies between the traditional operational means and emergency means. This document is intended for application to new nuclear power plants whose conceptual design is initiated after the publication of this document, but it may also be used for designing and implementing ERF in existing nuclear power plants or in any other nuclear facility. Detailed equipment design is outside the scope of this document. This document does not define the situations (reactor plant conditions, hazards and magnitudes of hazards) leading to mobilisation of emergency response teams and activation / use of the ERF. These aspects are usually addressed in the NPP Emergency Plan. However, the need for consistency of the ERF design and operation with the NPP Emergency Plan is within scope.
- Standard29 pagesEnglish languagesale 10% offe-Library read for1 day
This part of IEC/IEEE 62582 contains methods for condition monitoring of organic and polymeric materials in instrumentation and control cables using insulation resistance measurements in the detail necessary to produce accurate and reproducible results during simulated accident conditions. It includes the requirements for the measurement system and measurement procedure, and the reporting of the measurement results. NOTE Measurement of insulation resistance during simulated accident conditions with the aim of determining the lowest value during the accident in order to assess cable performance involves special requirements given in this document. Methods for measurement under stable (non-accident) conditions are available in other international standards, e.g. IEC 62631-3-3. The different parts of the IEC/IEEE 62582 series 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 qualification of equipment important to safety is found in IEC/IEEE 60780-323.
- Standard27 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 62855 provides the electrotechnical engineering guidelines for analysis of AC and DC electrical power systems in nuclear power plants (NPPs) in order to demonstrate that the power sources and the distribution systems have the capability for safe operation and shut down of the NPP, bringing it to a controlled state after an anticipated operational occurrence or accident conditions and finally reaching a safe state. The analytical studies discussed in this document provide assurance that the design bases are satisfied to meet their functional requirements under the conditions produced by the applicable design basis events. The studies provide assurance that the electrical power system is capable of supporting safety functions during all required plant conditions. NOTE The safety functions are described in IAEA Specific Safety Requirements SSR-2/1 related to the design of the nuclear power plants.. Analytical studies validate the robustness and adequacy of design margins and demonstrate the capability of electrical power systems to support plant operation for normal, abnormal, degraded and accident conditions. The analyses are used to verify that the electrical power system can withstand minor disturbances and that the consequences of major disturbances or failures do not degrade the capability of the electrical power systems to support safe shutdown of the plant and maintain the plant in shutdown condition. The analyses are performed with one or more of • simulation tools (software and hardware) that have been verified and validated, • hand calculations, and • tests. This document provides guidance on the types of analyses required to demonstrate that the plant's auxiliary power system can perform the required safety functions. This document does not provide specific details on how the analysis should be conducted. This document does not cover digital controllers (such as controllers for rectifiers, inverters, sequencers and electrical protection devices) used in electrical power systems. IEC 61513 gives recommendations that apply to the electronic controls and protective elements of the electrical power systems. This document does not include environmental conditions (i.e. temperature, humidity, etc.) or external events (seismic, flooding, fire, high energy electromagnetic pulse, etc.) that may impact equipment sizing or protection requirements. The external events lightning and geomagnetic storms are included. This document does not cover additional or unique requirements for stand-alone power system, such as power supplies for security measures in NPPs. Pertinent clauses of this document may be used as a guideline for such systems. Redundancy in the power system design can increase the availability of electrical power to critical plant equipment. Performing a probabilistic risk assessment (PRA) is a method of assessing system availability and optimizing design for high reliability. This document does not cover improving the reliability of NPP electrical power systems using statistical or diverse and redundant schemes. Requirements for safeguards of personnel involved with installation, maintenance and operation of electrical systems and general personal safety are outside the scope of this document. General guidance for lightning protection of equipment is provided in relevant clauses of this document. This document is intended to be used: • for verification of the design of new nuclear power plants, • for demonstrating the adequacy and impact of major modifications of electrical power systems in operating nuclear power plants, and • where there is a requirement to assess and establish operating limits and constraints for existing plants. Pertinent parts of this document can be used as guidance for decommissioning stages.
- Standard60 pagesEnglish languagesale 10% offe-Library read for1 day
This document establishes, for nuclear power plants2, a method of assignment of the functions specified for the plant into categories according to their importance to safety. Subsequent classification of the I&C and electrical power systems performing or supporting these functions, based on the assigned category, then determines relevant design criteria. The design criteria, when applied, ensure the achievement of each function in accordance to its importance to safety. In this document, the criteria are those of functionality, reliability, performance, environmental qualification (e.g. seismic) and quality assurance (QA). This document is applicable to: - the functions important to safety that are performed by I&C systems and supported by electrical power systems (categorization of I&C functions), - the I&C systems that enable those functions to be implemented (classification of I&C systems), - the electrical power systems that support those functions (classification of electrical power systems). The systems under consideration provide automated protection, closed or open loop control, information to the operating staff, and electrical power supply to systems. These systems keep the NPP conditions inside the safe operating envelope and provide automatic actions, or enable manual actions, that prevent or mitigate accidents, or that prevent or minimize radioactive releases to the site or wider environment. The I&C and electrical power systems that fulfil these roles safeguard the health and safety of the NPP operators and the public. This document follows the general principles given in IAEA Safety Requirement SSR-2/1 and Safety Guides SSG-30, SSG-34 and SSG-39, and it defines a structured method of applying the guidance contained in those codes and standards to the I&C and electrical power systems that perform functions important to safety in a NPP. This document is read in association with the IAEA guides together with IEC 61513 and IEC 63046 in implementing the requirements of the IEC 61508 series. The overall classification scheme of structures, systems and components for NPPs can be summarized as follows by Figure 1.
- Standard43 pagesEnglish languagesale 10% offe-Library read for1 day
This document:
• provides requirements and recommendations for the overall Electrical Power System. In
particular, it covers interruptible and uninterruptible Electrical Power Systems including
the systems supplying the I&C systems;
• is consistent and coherent with IEC 61513. Like IEC 61513, this document also highlights
the need for complete and precise requirements, derived from the plant safety goals.
Those requirements are prerequisites for generating the comprehensive requirements for
the overall Electrical Power System architecture, and for the electrical power supply subsystems;
• has to be considered in conjunction with and at the same level as IEC 61513. These two
standards provide a complete framework establishing general requirements for
instrumentation, control, and Electrical Power System for Nuclear Power Plants.
This document establishes:
• the high level specification and requirement to implement a suitable Electrical Power
System in a NPP that supports reactor systems important to safety. It also enables
electrical energy production providing the transmission grid with active and reactive power
and electro-mechanical inertia;
• the relationships between:
– the plant safety requirements and the architecture of the overall Electrical Power
System and its sub-systems (see Figure 1) including:
a) the contribution to the plant Defence in Depth;
b) the independency and redundancy provisions;
– the electrical requirements and the architecture of the Electrical Power System and its
sub-systems;
– the functional requirements and the architecture of the Electrical Power System and its
sub-systems;
– the requirements associated with the maintenance strategy and the architecture of the
Electrical Power System and its sub-systems;
• the design of Electrical power sub-systems (e.g. interruptible and uninterruptible);
• the requirements for supporting systems of Electrical Power System (HVAC, I&C, etc.);
• the Electrical Power System life-cycle framework.
This document does not cover the specification of:
• I&C systems;
• the transmission lines connecting to substations outside the NPP;
• electrical equipment requirements already defined in the industrial IEC standards;
• electrical power for security systems (e.g., fences, surveillance systems, entrance
control);
• lighting and socket facility.
This document does not consider power production requirements.
- Standard91 pagesEnglish languagesale 10% offe-Library read for1 day
See the scope of IEC 63046:2020. Adoption of IEC 63046:2020 is to be done without modification.
- Standard91 pagesEnglish languagesale 10% offe-Library read for1 day
This International Standard describes methods for establishing seismic qualification procedures
that will yield quantitative data to demonstrate that the equipment can meet its performance
requirements. This document is applicable to electrical, mechanical, instrumentation and control
equipment/components that are used in nuclear facilities. This document provides methods and
documentation requirements for seismic qualification of equipment to verify the equipment’s
ability to perform its specified performance requirements during and/or after specified seismic
demands. This document does not specify seismic demand or performance requirements. Other
aspects, relating to quality assurance, selection of equipment, and design and modification of
systems, are not part of this document. As seismic qualification is only a part of equipment
qualification, this document is used in conjunction with IEC/IEEE 60780-323.
The seismic qualification demonstrates equipment’s ability to perform its safety function(s)
during and/or after the time it is subjected to the forces resulting from at least one safe shutdown
earthquake (SSE/S2). This ability is demonstrated by taking into account, prior to the SSE/S2,
the ageing of equipment and the postulated occurrences of a given number of lower intensity
operating basis earthquake (OBE/S1). Ageing phenomena to be considered, if specified in the
design specification, are those which could increase the vulnerability of equipment to vibrations
caused by an SSE/S2.
- Standard85 pagesEnglish languagesale 10% offe-Library read for1 day
IEC 61468:2021 applies to in-core neutron detectors, viz. self-powered neutron detectors (SPNDs), which are intended for application in systems important for nuclear reactor safety: protection, instrumentation and control. This document contains SPND characteristics and test methods. In this document, the main sources of errors, and the possibilities for their minimization are also considered. This document contains requirements, recommendations and instructions concerning selection of SPND type and characteristics for various possible applications.
This document about SPNDs uses the basic requirements of IEC 61513 and IEC 60568 and complements them with more specific provisions in compliance with IAEA Safety Guides.
This second edition cancels and replaces the first edition, published in 2000, and its Amendment 1, published in 2003.
This edition includes the following significant technical changes with respect to the previous edition:
a. Title modified.
b. Justify the requirements for SPND characteristics in terms of influencing factors.
c. Align the terminology with the current state of the regulatory framework.
- Standard35 pagesEnglish languagesale 15% off