SiC fuel cladding: Macroscopic effects of radiation on mechanical and thermal properties from microstructural-scale characterisation and modelling

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Silicon carbide is a candidate material for accident tolerant fuel cladding applications in nuclear power reactors due to its good neutronic performance for light water reactors, high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to conventional Zircaloy fuel cladding under accident conditions. The development and investigation of these materials is particularly important in the light of the Fukushima event and the subsequent emphasis on Accident Tolerant Fuel concepts (ATF). Recent efforts have attempted to model the behaviour of several SiC-based cladding designs under realistic LWR operating conditions, and give predictions of stresses and failure probabilities throughout the fuel life. These models treat the very complex, highly structured SiC-SiC composite as a homogenous material. This assumption seriously limits the applicability and accuracy of these models and such models cannot be used effectively to optimize the cladding macro and microstructure.In order to design these materials to produce the best possible performance, a more refined model is needed representing accurately the composite microstructure and the behaviour of the various components For such a comprehensive model, the material properties of the constituent components of the composites, their mutual interfaces, and the way these respond to irradiation, are needed as input parameters.This project aims to develop localized materials assessment techniques to directly measure mechanical and thermal properties of the individual constituents of SiC-based claddings at the relevant micro-scale, before and after irradiation, in order to provide input parameters for a comprehensive model which will be developed and validated in this program.