Abstract
In the water pressurised nuclear reactors, the fuel rod cladding is the first barrier against radioactive isotopes release. Its integrity must be demonstrated all along the fuel rod irradiation with increasing strenuous operating conditions. This paper deals with a study made with Electricité de France (EDF) in order to improve the understanding and modelling of the thermomechanical behaviour of fuel rods under these more arduous conditions. The aim of this study is to evaluate the separate influences of structural and material parameters variability on Pellet–Cladding Interaction (PCI). The following parameters have been tested by 3D simulations for a common ramping condition: axial and radial pellet cracks numbers, pellet fragment size, relative fragments displacement, non-symmetrical configuration, pellet–pellet friction and pellet–cladding friction. The second part of the article deals with the development of a model, which gives a better assessment of cladding stress concentration near radial fuel cracks. The implementation in the 1D fuel rod EDF code CYRANO3 and the validation of the model are presented.
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