| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 7963223 | Journal of Nuclear Materials | 2018 | 41 Pages |
Abstract
Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20â¯MeV Kr8+ ions at 400â¯Â°C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)2 Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.
Keywords
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Physical Sciences and Engineering
Energy
Nuclear Energy and Engineering
Authors
A. Wu, J. Ribis, J.-C. Brachet, E. Clouet, F. Leprêtre, E. Bordas, B. Arnal,