International audienceThe oxidation rate of Zircaloy-4 fuel cladding in nuclear reactor strongly increases at high burnups. This kinetics acceleration could be in large part due to the irradiation damage. The irradiation effect of the post-transition oxide layer on the Zircaloy-4 corrosion rate has been investigated using protons. As previously observed on pre-transition oxide layer, irradiation defects increases the oxidation rate of the alloy up to around 30 days in autoclave in agreement with the irradiation defect annealing characterized by Raman spectroscopy. The model proposed in previous works is able to describe the oxidation rate after irradiation of post-transition oxide layers
A one dimensional numerical model is presented to predict oxide scale growth and failure in zirconiu...
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering...
International audienceThe main purpose of this paper is to summarize our results obtained during res...
International audienceThe oxidation rate of Zircaloy-4 fuel cladding in nuclear reactor strongly inc...
International audienceA strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observe...
International audienceIrradiation damage in fuel cladding material is mainly caused by the neutron f...
International audienceThe oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressur...
International audienceThe corrosion process (oxidation and hydriding) of the zirconium alloy claddin...
[[abstract]]Specimens of Zircaloy-2 plate were irradiated with 2 MeV proton at 350 °C up to tota...
[[abstract]]In previous studies of irradiation effects on the nodular corrosion resistance of Zircal...
This thesis is part of the MUZIC-3 (Mechanistic Understanding of Zirconium Corrosion) project, with ...
The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel ...
The oxidation of zirconium alloys under aqueous conditions has been studied for more than 50 years ...
A detailed study has been carried out on recrystallised Zr-1.0Nb alloys corroded and irradiated unde...
A one dimensional numerical model is presented to predict oxide scale growth and failure in zirconiu...
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering...
International audienceThe main purpose of this paper is to summarize our results obtained during res...
International audienceThe oxidation rate of Zircaloy-4 fuel cladding in nuclear reactor strongly inc...
International audienceA strong increase of the oxidation rate of Zircaloy-4 fuel cladding is observe...
International audienceIrradiation damage in fuel cladding material is mainly caused by the neutron f...
International audienceThe oxidation kinetics acceleration observed in Zircaloy-4 cladding in Pressur...
International audienceThe corrosion process (oxidation and hydriding) of the zirconium alloy claddin...
[[abstract]]Specimens of Zircaloy-2 plate were irradiated with 2 MeV proton at 350 °C up to tota...
[[abstract]]In previous studies of irradiation effects on the nodular corrosion resistance of Zircal...
This thesis is part of the MUZIC-3 (Mechanistic Understanding of Zirconium Corrosion) project, with ...
The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel ...
The oxidation of zirconium alloys under aqueous conditions has been studied for more than 50 years ...
A detailed study has been carried out on recrystallised Zr-1.0Nb alloys corroded and irradiated unde...
A one dimensional numerical model is presented to predict oxide scale growth and failure in zirconiu...
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering...
International audienceThe main purpose of this paper is to summarize our results obtained during res...