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SiC-SiC ceramic matrix composites are candidate materials for fuel cladding in Generation IV nuclear fission reactors and as accident tolerant fuel clad in current generation plant. Experimental methods are needed that can detect and quantify the development of mechanical damage, to support modelling and qualification tests for these critical components. In situ observations of damage development have been obtained of tensile and C-ring mechanical test specimens of a braided nuclear grade SiC-SiC ceramic composite tube, using a combination of ex situ and in situ computed X-ray tomography observation and digital volume correlation analysis. The gradual development of damage by matrix cracking and also the influence of non-uniform loading are examined.
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- 2 Finished
1/10/12 → 30/09/15