ISO/TC 85/SC 6 - Reactor technology

This document specifies the utilization and characteristics of instrumentation used to detect seismic events at nuclear power plants with water cooled reactors. The document can also be applied to other nuclear facilities after verifying its applicability. The following types of electrical systems and equipment are not covered by this document: — seismic instrumentation involved in the implementation of nuclear safety functions as defined by IEC 61226, for example automatic shutdown systems; — seismic instrumentation not involved in the implementation of nuclear safety functions as defined by IEC 61226 but which due, for example, to close proximity to other safety classified systems, requires hardware qualification to be performed. Such systems are specified, designed, manufactured, qualified, operated and dismantled according to the relevant requirements of IEC standards, in particular IEC 61513 and the lower level IEC standards according to the safety class and technologies used. Seismic instrumentation used for the implementation of seismic reactor trip systems are developed according to the requirements of IEC 63186. An automatic shutdown system is not covered by this document. This document specifies the requirements to be fulfilled by the seismic instrumentation such that, firstly, it can be ascertained whether any of the design quantities on which the plant walk-down level and the inspection levels are based have been exceeded and that, secondly, the recording of the time history of the earthquake provides the necessary input values for a post-seismic analysis. The requirements are specified such that, independent of the detection and recording system, comparable results within tolerances are achieved in the time range as well as the frequency range.

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.

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.

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.

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.

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.

This document applies to the reactor physics tests that are performed following a refuelling or other core alteration of a PWR for which nuclear design calculations are required. This document does not address the physics test program for the initial core of a commercial PWR. This document specifies the minimum acceptable startup reactor physics test program to determine if the operating characteristics of the core are consistent with the design predictions, which provides assurance that the core can be operated as designed. This document does not address surveillance of reactor physics parameters during operation or other required tests such as mechanical tests of system components (for example, the rod drop time test), visual verification requirements for fuel assembly loading, or the calibration of instrumentation or control systems (even though these tests are an integral part of an overall program to ensure that the core behaves as designed).

This document provides guidance in the preparation, verification, and validation of group-averaged neutron and gamma-ray cross sections for the energy range and materials of importance in radiation protection and shielding calculations for nuclear reactors[1], see also Annex A. [1] This edition is based on ANSI/ANS-6.1.2-2013[1].

This document provides the basis for calculating the decay heat power of non-recycled nuclear fuel of light water reactors. For this purpose the following components are considered: — 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. This document applies to light water reactors (pressurized water and boiling water reactors) loaded with a nuclear fuel mixture consisting of 235U and 238U. Application of the fission product contribution to decay heat developed using this document to other thermal reactor designs, including heavy water reactors, is permissible provided that the other contributions from actinides and neutron capture are determined for the specific reactor type. Its application to recycled nuclear fuel, like mixed-oxide or reprocessed uranium, is not permissible. The calculation procedures apply to decay heat periods from 0 s to 109 s.

This document specifies an analytical method for determining heavy water isotopic purity by Fourier transform infrared spectroscopy (FTIR). It is applicable to the determination of the whole range of heavy water concentration. The method is devoted to process controls at the different steps of the process systems in heavy water reactor power plant or any other related areas. The method can be applied for heavy water isotopic purity measurements in a heavy water reactor power plant or research reactor, heavy water production factory and heavy water related areas.

This document specifies requirements for the ice plug technique with liquid nitrogen or dry ice as refrigerant (cryogenic medium) on metal pipes of nuclear power plants. The freezing liquid can be water or water mixture (e.g. boric acid mixture). This document specifies technical requirements of ice plug generation, formation judgment and removal, measures before, during and after ice plugging and requirements for personnel and non-destructive testing. The application of the ice plug isolation technique is principally not allowed on cladded pipes or pipes with internal coatings. The application for pressure test is not in the scope of this document and will be qualified separately.

This document specifies the basic requirements of thermal insulation design of reactor coolant system (RCS) equipment and piping. Among thermal insulation of various RCS equipment and piping, the following two kinds of thermal insulations are described in detailed based on common design logic and requirements: — thermal insulation of reactor pressure vessel (RPV); — thermal insulation of RCS piping and other equipment. This document is valid for two types of thermal insulation: — metallic thermal insulation; — non-metallic thermal insulation. This document mainly applies to nuclear power plants with pressurized water reactor (PWR). For other reactor types, this document can be taken as reference.

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.

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).

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.

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.

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.

The document provides: — guidelines for determining the thermal effects to consider on fire barriers inside a given room; — guidelines for determining the global performance of the fire barriers based on standard test characterization; — guidelines for assessing the need for additional tests to verify the robustness of the solution. Requirements of applicable standards, numerical tools validation and verification (V&V), and the expected qualification of fire resistance laboratories are detailed. The limitations of the method's applicability and scope are discussed. The purpose and justification of this document is to describe a new methodology for the verification of the efficiency of fire barriers, which is initially based on a standardized fire resistance test. The significance of this work relates to the fact that the present methodology will enhance the level of safety by providing more realism to hazards analysis in combination with standardized test data. It completes the standard ISO-fire rating required for justifying the performance. The most relevant benefit of this method concerns the determination of the global performance of a barrier in a fire of extended duration compared to the classification given by the ISO-fire rating.

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.

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.

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).

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.

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.

Identifies the typical parameters and terms necessary for collecting and exchanging maintainability data. The definitions are valid for corrective maintenance. For preventive maintenance, a similar classification can be drawn up. The general guidelines on the exchange of reliability data for nuclear power plants are given in ISO 6527.

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.

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.

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.

Specifies the requirements for designing and the data required and the way in which they should be used in order to determine the earth motions to be taken as the design basis earthquake (DBE), moreover the way in which the proof of seismic design adequacy should be established and documented. Determination of DBE is covered by both probabilistic and deterministic methods. Is entirely applicable when the DBE is greater than or equal to intensity VII on the MSK scale. When the DBE is less, the structural analysis could also be performed by using simpler rules.

The output of a data collection system is strongly dependent on the quality of the information collected. Before starting such a system, it is necessary to clearly define the following items: the overall goal, the suppliers of field data, the users of processed data, the terms and expressions used, the means used to collect data and to treat them, the questions to be answered by field data, field data needed. The standard gives a comprehensive guidance to ensure quality of availability and reliability data collected in nuclear power plants.

Identifies the typical parameters of a component that permit it to be characterized unequivocally and to allow the corresponding reliability data to be associated with those of other components having equivalent typical parameters. Parameters refer to technical characteristics including the physical principle of operation and quality level and to actual operating conditions and maintenance and test intervals. Data may be represented both in a historical and in a statistical form.