Effect of pre-oxide on Zircaloy-4 high temperature steam oxidation

Breakaway oxidation of zirconium alloys exposed to steam around 1000 °C

(2020) M. Le Saux et al.   CORROSION SCIENCE

Effect of a pre-oxide on the high temperature steam oxidation of Zircaloy-4 and M5Framatome alloys

(2019) M. Le Saux et al.   JOURNAL OF NUCLEAR MATERIALS

Determination of dose rate effects on Zircaloy oxidation using proton irradiation and oxygen transport modeling

(2019) Michael Reyes et al.   JOURNAL OF NUCLEAR MATERIALS

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

(2019) Takafumi Narukawa et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

The correlation between tetragonal phase and the undulated metal/oxide interface in the oxide films of zirconium alloys

(2019) Jingjing Liao et al.   JOURNAL OF NUCLEAR MATERIALS

Experimental study and numerical simulation of high temperature (1100–1250 °C) oxidation of prior-oxidized zirconium alloy

(2016) Benoît Mazères et al.   CORROSION SCIENCE

The microstructure and microtexture of zirconium oxide films studied by transmission electron backscatter diffraction and automated crystal orientation mapping with transmission electron microscopy

(2014) A. Garner et al.   ACTA MATERIALIA

Behavior of High Burn-up Fuel Cladding under LOCA Conditions

(2012) Fumihisa NAGASE et al.   JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

HIGH TEMPERATURE OXIDATION OF NB-CONTAINING ZR ALLOY CLADDING IN LOCA CONDITIONS

(2012) Toshinori Chuto et al.   Nuclear Engineering and Technology

Comments to papers of J.H. Kim et al. [1] and M. Große et al. [2] recently published in JNM “On the hydrogen uptake of Zircaloy-4 and M5™ alloys subjected to steam oxidation in the 1100–1250°C temperature range”

(2009) J.-C. Brachet et al.   JOURNAL OF NUCLEAR MATERIALS

Ductile-to-brittle transition of oxidised Zircaloy-4 and E110 claddings

(2007) Zoltán Hózer et al.   JOURNAL OF NUCLEAR MATERIALS