Abstract The Fukushima-Daiichi accident revealed that the zirconium fuel claddings have
the significant safety risk of hydrogen detonation due to the strong oxidation and hydrogen …

Zirconium-based alloys have served the nuclear industry for several decades due to their
acceptable properties for nuclear cores of light water reactors (LWRs). However, severe …

The fretting corrosion behaviors and the associated mechanisms of Zr alloy claddings with
and without Cr coating in the primary water (310° C/15.5 MPa) of the simulated pressurized …

The oxidation of chromium-coated zirconium-based alloys is studied under steam at
temperatures ranging from 800° C up to 1500° C and for oxidation times ranging from a few …

The isothermal oxidation behavior of the Cr-coated Zircaloy-4 at 1200° C in steam is
comprehensively investigated. Oxidation kinetics transition is observed when outward …

Cr-based coatings with or without a Mo interlayer were deposited on a Zr substrate by multi-
arc ion plating, and the effect of the Mo interlayer on the oxidation behavior of the coatings at …

Abstract The Cr-Zr interaction of two types of Cr coated Zr alloy accident tolerant fuel (ATF)
claddings, deposited by cold spraying and magnetron sputtering, was studied in argon in the …

Cr coatings with the thickness of 4.5–9.0 μm and dense/columnar microstructure were
deposited onto Zr alloy by cooled or hot target magnetron sputtering. Steam oxidation tests …

Cr coatings have been deposited on the outer surface of zircaloy-4 tube with well control of
thickness variations along the tube axis. A series of samples have been oxidized in high …

The isothermal steam oxidation behavior of chromium-coated zirconium alloy cladding at
1200℃ is studied. The cladding substrate is Zr-1 Nb alloy with outer diameter of 9.5 mm and …