Oxide dispersion strengthened (ODS) ferritic steels are considered promising candidates as cladding tubes for Generation IV nuclear reactors. In such reactors, irradiation damage can reach more than 150 dpa at temperatures ranging from 400 to 650 °C. Thus nanoparticle stability has to be guaranteed in order to ensure that these materials possess excellent creep properties. Using Fe ions, ODS steels were irradiated at 500 °C up to 150 dpa. At this temperature the nano-oxide population evolution under irradiation is similar to that observed after annealing at high temperature. It consists of a slight increase in the particle size and a slight decrease in the density, which can be both explained by an Ostwald ripening mechanism. Conversely, irradiations performed at room temperature using Au ions lead to a complete dissolution of the oxide particles, in agreement with the estimation of ballistic vs. radiation enhanced diffusion effects

Raman spectroscopy is an efficient technique for studying the evolution of microstructure of materials under irradiation. For that purpose, a Raman spectrometer has been recently installed at the JANNUS‐Saclay platform. In this paper, we describe the new setup for in situ experiments. These in situ experiments allowed following the microstructural evolution of different materials (SiC, ZrO2 and B4C) as a function of ion fluence on a single sample (either single crystal or polycrystalline ceramics) under the same irradiation conditions. Our results show that Raman spectroscopy is a versatile non‐contact technique for studying on‐line crystalline phase changes or amorphization of irradiated iono‐covalent solids. A detailed analysis of Raman spectra is provided for the three materials (SiC, ZrO2 and B4C) investigated in this study, revealing quite different behaviors upon irradiation. Basically, Raman spectroscopy gives insight on these evolutions at the level of bonds given by specific phonon modes, in good agreement with Rutherford backscattering channeling (RBS/C), X‐ray diffraction (XRD) or transmission electron microscopy (TEM) data, which provide information at a long‐range scale. Copyright © 2015 John Wiley & Sons, Ltd.

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 20 MeV Kr 8þ ions at 400 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.