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.

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IEC/IEEE 60980-344:2020 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.

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IEC 61031:2020 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.

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

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

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

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

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

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ISO 19226:2017 provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity, between the ends of active fuel assemblies, given the neutron source in the core.
NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production.
The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs).
ISO 19226:2017 also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

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ISO 19226:2017 provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity, between the ends of active fuel assemblies, given the neutron source in the core.
NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production.
The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs).
ISO 19226:2017 also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

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This document specifies requirements for the unique identification of fuel assemblies utilized in nuclear power plants. It was developed primarily for commercial light-water reactor fuel, but can be used for any reactor fuel contained in discrete fuel assemblies that can be identified with an identification code as specified by this document. This document defines the characters and proposed sequence to be used in assigning identification to the fuel assemblies. The identification is intended to be borne by the fuel assembly throughout its lifetime. This document aims at providing an organizing principle for fuel assembly identification systems in order to guarantee unequivocal identification at any time and any place in the world (see also IAEA Safety Guide No. GS-G-3.5). Considering that existing standards for fuel assembly identification (such as ANSI/ANS-57.8-1995, DIN 25433, IAEA Safety Guide No. GS-G-3.5) ensure unequivocal identification in their respective fields of application, this document allows without restriction the further application of these standards. Moreover, it is intended that this document be used as a guideline for new definitions of identification systems.

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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.

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ISO 18077:2018 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[1]. ISO 18077:2018 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). ISO 18077:2018 assumes that the same previously accepted analytical methods are used for both the design of the reactor core and the startup test predictions. It also assumes that the expected operation of the core will fall within the historical database established for the plant and/or sister plants. When major changes are made in the core design, the test program should be reviewed to determine if more extensive testing is needed. Typical changes that might fall in this category include the initial use of novel fuel cycle designs, significant changes in fuel enrichments, fuel assembly design changes, burnable absorber design changes, and cores resulting from unplanned short cycles. Changes such as these may lead to operation in regions outside of the plant's experience database and therefore may necessitate expanding the test program. [1] The good practices discussed in this document should be considered for use in the physics test program for the initial core of a commercial PWR. One test that provides useful information (without additional test time) is the hot-zero-power to hot-full-power reactivity measurement.

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ISO 18075:2018 provides guidance for performing and validating the sequence of steady-state calculations leading to prediction, in all types of operating UO2-fuel commercial nuclear reactors, of: - reaction-rate spatial distributions; - reactivity; - change of nuclide compositions with time. ISO 18075:2018 provides: a) guidance for the selection of computational methods; b) criteria for verification and validation of calculation methods used by reactor core analysts; c) criteria for evaluation of accuracy and range of applicability of data and methods; d) requirements for documentation of the preceding.

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ISO 18229:2018 defines the essential technical requirements that are addressed in the process of design and construction of Generation IV (GEN IV) nuclear reactors. It does not address operation, maintenance and in-service inspection of reactors. Six reactor concepts are considered for GEN IV: the sodium fast reactor, the lead fast reactor, the gas fast reactor, the very high temperature reactor, the supercritical water reactor and the molten salt reactor. Annex A details the main characteristics for the different concepts. The scope of application of this document is limited to mechanical components related to nuclear safety and to the prevention of the release of radioactive materials ? that are considered to be important in terms of nuclear safety and operability, ? that play a role in ensuring leaktightness, partitioning, guiding, securing and supporting, and ? that contain and/or are in contact with fluids (such as vessels, pumps, valves, pipes, bellows, box structures, heat exchangers, handling and driving mechanisms).

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ISO 19226:2017 provides a procedure for the evaluation of irradiation data in the region between the reactor core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity, between the ends of active fuel assemblies, given the neutron source in the core. NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production. The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water reactors (BWRs), and pressurized heavy water reactors (PHWRs). ISO 19226:2017 also provides a procedure for evaluating neutron damage properties at the reactor pressure vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum. Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated number of atomic displacements are important data to be used for reactor life management.

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IEC 62516-2:2011 specifies the characteristics and requirements for interactive data services using binary format for scene (BIFS) in the terrestrial digital multimedia broadcasting (T-DMB) receiver.

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ISO 26802:2010 specifies the applicable requirements related to the design and the operation of containment and ventilation systems of nuclear power plants and research reactors taking into account the following. For nuclear power plants, ISO 26802:2010 addresses only reactors that have a secondary confinement system based on IAEA recommendations. For research reactors, ISO 26802:2010 applies specifically to reactors for which accidental situations can challenge the integrity or leak-tightness of the containment barrier, i.e. in which a high-pressure or -temperature transient can occur and for which the isolation of the containment building and the shut-off of the associated ventilation systems of the containment building is required. The requirements of ISO 26802:2010 apply to research reactors in which the increase of pressure or temperature during accidental situations do not risk damaging the ventilation systems, although the requirements applicable for the design and the use of ventilation systems are given in ISO 17873.

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The scope of IEC 61008-1 and IEC 61009-1 applies with the following additions. IEC 62423:2009 specifies requirements and tests for Type F and Type B RCDs (Residual Current Devices). Requirements and tests given in this standard are in addition to the requirements of Type A residual current devices. This standard can only be used together with IEC 61008-1 and IEC 61009-1. Type F RCCBs (Residual Current Circuit Breaker) and Type F RCBOs (Residual current Circuit Breaker with Overcurrent protection) with rated frequency 50 Hz or 60 Hz are intended for installations when frequency inverters are supplied between phase and neutral or phase and earthed middle conductor and are able to provide protection in case of alternating residual sinusoidal at the rated frequency, pulsating direct residual currents and composite residual currents that may occur. Type B RCCBs and Type B RCBOs are able to provide protection in case of alternating residual sinusoidal currents up to 1 000 Hz, pulsating direct residual currents and smooth direct residual currents. This second edition cancels and replaces the first edition published in 2007 and constitutes a technical revision. The main changes from the first edition are as follows:
- requirements and tests for Type F RCD have been introduced;
- requirements and test for two-pole Type B RCD have been introduced;
- new additional requirements and tests for Type B RCDs have been introduced to cover requirements for Type F too. The contents of the corrigendum of December 2011 have been included in this copy.

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IEC 60768:2009 provides criteria for the design, selection, testing, calibration and functional location of equipment for the monitoring of radioactive substances within plant-process streams during normal operation conditions and anticipated operational occurrences. Is only applicable to continuous in-line or on-line measurement. This new edition includes the following significant technical changes with respect to the previous edition:
- it clarifies the definitions;
- up-dates the reference to new standards published since the first issue;
- updates the units of radiation.
This publication is to be read in conjunction with IEC 60951-1:2009, IEC 60951-2:2009, IEC 60951-3:2009 and IEC 60951-4:2009.

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IEC 61000-4-28:1999+A1:2001+A2:2009 establishes a reference for evaluating the immunity of electric and electronic equipment when subjected to variations of the power frequency. Only conducted phenomena are considered, including immunity tests for equipment connected to public and industrial networks. This consolidated version consists of the first edition (1999), its amendment 1 (2001) and its amendment 2 (2009). Therefore, no need to order amendments in addition to this publication.

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IEC/TR 62096:2009 is intended to support owners of a nuclear power plant in the decision-making process and in the preparation for partial or complete modernization of the I&C. For this, it provides a summary of the motivating factors for I&C modernization, the principal options for the elaboration of different scenarios for I&C modernization, the technical and economic criteria for the selection of a long term I&C strategy, the principal aspects to be taken into account for a detailed technical feasibility study. This new edition includes the following significant technical changes with respect to the previous edition:
- update on references, taking into account standards published since the previous edition;
- update on the terminology, incorporation of a number of clarifications proposed by National Committees.

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Specifies methods for the determination of mass gain and for the surface inspection of products of zirconium and its alloys when corrosion tested in water at 360 °C or in steam at or above 400 °C. Applicable to wrought products, castings and powder metallurgical products.

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See the scope of IEC/IEEE 60980-344:2020. Adoption of IEC/IEEE 60980-344:2020is to be done without modification.

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Is applicable to capacitors according to IEC 60110-1 and gives the requirements for the ageing test, destruction test and requirements for disconnecting internal fuses for these capacitors.

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Is applicable both to indoor capacitor units and indoor capacitor banks intended to be used, particularly, for power factor correction in induction heating, melting, stirring or casting installations, and similar applications with controlled or adjustable a.c. voltage systems in a frequency range up to 50 kHz, and with a rated voltage not exceeding 3,6 kV.

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Deals with the standard conditions and methods of measurement on high definition television (HDTV) displays. Deals with the determination of performance and permits comparision of equipment by listing characteristics useful for specifications and laying down uniform methods of measurement.

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Specifies methods for the determination of mass gain and for the surface inspection of products of zirconium and its alloys when corrosion tested in water at 360 °C or in steam at or above 400 °C. Applicable to wrought products, castings and powder metallurgical products.

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Provides the basis for calculating the decay heat power of non-recycled nuclear fuel considering: the contribution of the fission products from nuclear fission; the contribution of the actinides; the contribution of isotopes resulting from neutron capture in fission products.

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Is applicable to all operations with plutonium, uranium 233, uranium enriched in the 235 isotope, and other fissionable materials. Does not require separate additional instrumentation when the operating instrumentation of the facilities meets its requirements. Is principally concerned with gamma-radiation rate-sensing systems. Annex A refers to the specification of a minimum accident of concern, annex B provides examples of application, and annex C provides guidance for development of emergency plans. Does not include details of administrative steps or specific design and description of instrumentation.

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This document defines the essential technical requirements that are addressed in the process of design and construction of Generation IV (GEN IV) nuclear reactors. It does not address operation, maintenance and in-service inspection of reactors.
Six reactor concepts are considered for GEN IV: the sodium fast reactor, the lead fast reactor, the gas fast reactor, the very high temperature reactor, the supercritical water reactor and the molten salt reactor.
Annex A details the main characteristics for the different concepts.
The scope of application of this document is limited to mechanical components related to nuclear safety and to the prevention of the release of radioactive materials
—   that are considered to be important in terms of nuclear safety and operability,
—   that play a role in ensuring leaktightness, partitioning, guiding, securing and supporting, and
—   that contain and/or are in contact with fluids (such as vessels, pumps, valves, pipes, bellows, box structures, heat exchangers, handling and driving mechanisms).

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This document defines the essential technical requirements that are addressed in the process of design and construction of Generation IV (GEN IV) nuclear reactors. It does not address operation, maintenance and in-service inspection of reactors.
Six reactor concepts are considered for GEN IV: the sodium fast reactor, the lead fast reactor, the gas fast reactor, the very high temperature reactor, the supercritical water reactor and the molten salt reactor.
Annex A details the main characteristics for the different concepts.
The scope of application of this document is limited to mechanical components related to nuclear safety and to the prevention of the release of radioactive materials
—   that are considered to be important in terms of nuclear safety and operability,
—   that play a role in ensuring leaktightness, partitioning, guiding, securing and supporting, and
—   that contain and/or are in contact with fluids (such as vessels, pumps, valves, pipes, bellows, box structures, heat exchangers, handling and driving mechanisms).

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Applies to equipment for the monitoring of radioactive substances within plant process streams of stationary nuclear power plants with light-water reactors during specified normal operation (routine operation) and during anticipated operational occurrences (incidents). Provides criteria for the design, selection, functional location, testing and calibration of stationary radiation equipment to be used for continuous monitoring of plant process streams.

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Specifies requirements for the unique identification. Was developed primarily for commercial light-water reactor fuel, but may be used for any reactor fuel contained in discrete fuel assemblies that can be identified with an identification code. Defines the characters and proposed sequence to be used in assigning identification to the fuel assemblies. The identification is intended to be borne by the fuel assembly throughout its lifetime. Aims at providing an organizing principle for fuel assembly identification systems in order to guarantee unequivocal identification at any time and any place in the world.

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