Abstract
The fuel cladding is one of the most important structural components for maintaining the integrity of a fuel channel and for safely exploitation of a nuclear power plant. The corrosion behavior of a fuel cladding material, Zy-4, under high pressure and temperatures conditions, was analyzed in a static isothermal autoclave under simulated primary water conditions—a LiOH solution at 310 °C and 10 MPa for up to 3024 h. After this, the oxides grown on the Zy-4 sample surface were characterized using electrochemical measurements, gravimetric analysis, metallographic analysis, SEM and XPS. The maximum oxide thicknesses evaluated by gravimetric and SEM measurements were in good agreement; both values were around 1.2 µm. The optical light microscopy (OLM) investigations identified the presence of small hydrides uniformly distributed horizontally across the alloy. EIS impedance spectra showed an increase in the oxide impedance for the samples oxidized for a long time. EIS plots has the best fit with an equivalent circuit which illustrated an oxide model that has two oxide layers: an inner oxide layer and outer layer. The EIS results showed that the inner layer was a barrier layer, and the outer layer was a porous layer. Potentiodynamic polarization results demonstrated superior corrosion resistance of the samples tested for longer periods of time. By XPS measurements we identified all five oxidation states of zirconium: Zr0 located at 178.5 eV; Zr4+ at 182.8 eV; and the three suboxides, Zr+, Zr2+ and Zr3+ at 179.7, 180.8 and 181.8 eV, respectively. The determination of Vickers microhardness completed the investigation.
Highlights
For materials used in the sustainable energy domain, investigations of properties are continuous, to enhance selection [1]
The corrosion processes of all metallic materials have determining roles in efficiently and safely operating a nuclear power plant (NPP), and their oxidation processes are largely investigated as functions of the environment for their industrial exploitation [2,3,4,5]
The objective of this work is to contribute to knowledge about future materials for nuclear power plants— [26], to describe the corrosion behavior of a fuel cladding material, Zircaloy-4, under high pressure and temperatures conditions
Summary
For materials used in the sustainable energy domain, investigations of properties are continuous, to enhance selection [1]. Zirconium-based alloys serve the nuclear industry as metallic tubes called cladding, where the nuclear fuel element is located and assembled. They are exposed to an aggressive mediums of water at high pressures and high temperatures Ppm Li), in addition to high fluxes of energetic neutrons In this environment, the mechanical strength and corrosion behavior of the materials are of extreme importance. Being used as fuel cladding materials in pressurized heavy water reactors (PHWR), Zr alloys represent the first containment barrier to fission products. Their mechanical integrity has to be ensured during their life cycles
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