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dc.contributor.authorGARRIDO Oriol Costaen_GB
dc.contributor.authorCIZELJ Leonen_GB
dc.contributor.authorSIMONOVSKI IGORen_GB
dc.date.accessioned2012-07-27T00:01:06Z-
dc.date.available2012-07-26en_GB
dc.date.available2012-07-27T00:01:06Z-
dc.date.created2012-07-25en_GB
dc.date.issued2012en_GB
dc.date.submitted2011-09-19en_GB
dc.identifier.citationNUCLEAR ENGINEERING AND DESIGN vol. 246 no. May 2012 p. 115-122en_GB
dc.identifier.issn0029-5493en_GB
dc.identifier.urihttp://www.sciencedirect.com/science/article/pii/S0029549311005218en_GB
dc.identifier.urihttp://publications.jrc.ec.europa.eu/repository/handle/JRC66779-
dc.description.abstractThe extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g. pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR reactor coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.en_GB
dc.description.sponsorshipJRC.F.5-Nuclear Reactor Safety Assessmenten_GB
dc.format.mediumPrinteden_GB
dc.languageENGen_GB
dc.publisherELSEVIER SCIENCE SAen_GB
dc.relation.ispartofseriesJRC66779en_GB
dc.titleModular 3-D solid finite element model for fatigue analyses of a PWR coolant systemen_GB
dc.typeArticles in periodicals and booksen_GB
dc.identifier.doi10.1016/j.nucengdes.2011.06.038en_GB
JRC Directorate:Energy, Transport and Climate

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