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dc.contributor.authorNILSSON Karl-Fredriken_GB
dc.contributor.authorPAFFUMI Elenaen_GB
dc.contributor.authorRADU Vasileen_GB
dc.date.accessioned2013-01-10T01:02:07Z-
dc.date.available2013-01-09en_GB
dc.date.available2013-01-10T01:02:07Z-
dc.date.created2012-12-18en_GB
dc.date.issued2012en_GB
dc.date.submitted2012-12-05en_GB
dc.identifier.citationBook of abstracts - Fracture Mechanics for Durability, Reliability and Safety - 19th European Conference on Fracture p. 399en_GB
dc.identifier.isbn978-5-905576-18-8en_GB
dc.identifier.urihttp://publications.jrc.ec.europa.eu/repository/handle/JRC77346-
dc.description.abstractThe paper describes a five-step procedure for thermal fatigue for steel pipes in light-water reactors (LWR) with a maximum temperature of 350° C that has been developed in European NESC and NULIFE projects. The first step is a simple screening criterion based on operational experience. The four more advanced steps address crack initiation or crack propagation using a simplified “sinusoidal” procedure or spectrum loading for the loads. The crack initiation and propagation are in all cases described by fatigue curves and Paris law respectively. In the sinusoidal procedure the thermal fluctuation have a sinusoidal variation with a constant frequency and where the crack initiation and propagation life are determined from the frequency (typically 0.1 – 1 Hz) that gives the shortest life measured in time. The procedure will be illustrated by comparison with real cases. The sinusoidal approach is generally conservative but not unreasonably conservative since the frequency of the load spectra is similar to the critical frequency. The paper will also illustrate and quantify the importance of different key parameters such as heat transfer coefficient, shape of the thermal load (sinusoidal vs. squared), axial load and crack shape. We will also show how the procedure can be used to determine the life of a pipe component with a given load, or how to determine the allowable load for a prescribed design life. Future reactors such as Sodium Fast Reactors (SFR) will have higher temperatures (550° C) for which creep effects need to be taken into account. To this end we are working on how to extend the procedure to account for creep-fatigue and we will present some preliminary ideas.en_GB
dc.description.sponsorshipJRC.F.4-Nuclear Reactor Integrity Assessment and Knowledge Managementen_GB
dc.format.mediumPrinteden_GB
dc.languageENGen_GB
dc.publisherEuropean Structure Integrity Society (ESIS)en_GB
dc.relation.ispartofseriesJRC77346en_GB
dc.titleEngineering Procedure for Thermal Fatigue in Nuclear Componentsen_GB
dc.typeArticles in periodicals and booksen_GB
JRC Directorate:Energy, Transport and Climate

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