On-going studies at CEA on chromium coated zirconium based nuclear fuel claddings for enhanced accident tolerant LWRS fuel

International audienceChromium coatings aiming to protect the current zirconium alloy nuclear fuel cladding materials from high temperature steam oxidation, especially in accidental conditions, are being developed at CEA in the framework of the French Nuclear Institute in partnership with AREVA and EDF. The present paper focuses on recent results obtained on chromium-coated cladding on Zircaloy-4 and substrate. A complementary paper is presented by AREVA. Typical as-fabricated microstructures of chromium coatings are illustrated. Preliminary autoclave oxidation tests at 415°C (steam, 100 bars) have been performed on both uncoated reference materials and on chromium coated Zircaloy-4 based samples, showing very encouraging results for the last generation of Cr coatings fabricated. Additionally, results obtained after High Temperature (HT) oxidation in steam are presented. It is shown that the chromium coatings developed so far may significantly improve the post-HT oxidation clad mechanical properties (i.e. ductility and strength) compared to conventional uncoated materials. As a consequence, the developed chromium coatings provide significant additional margins for loss-of-coolant accident (LOCA) and, to some extent, for beyond LOCA conditions