Abstract
The manuscript is focused on corrosion behavior of a Cr coating under CANada Deuterium Uranium(CANDU) primary circuit conditions. The Cr coating is obtained via the thermionic vacuum arc procedure on Zircaloy -4 cladding. The surface coating characterization was performed using metallographic analysis and scanning electron microscopy (SEM) with an energy dispersive spectra detector (EDS), X-ray diffraction (XRD), and X-ray photoelectron spectroscopy (XPS) investigations. The thickness of the Cr coating determined from SEM images is around 500 nm layers After the autoclaving period, the thickness of the samples increased in time slowly. The kinetic of oxidation established a logarithmic oxidation law. The corrosion tests for various autoclaving periods of time include electrochemical impedance spectroscopy (EIS) and potentiodynamic tests, permitting computing porosity and efficiency of protection. All surface investigations sustain electrochemical results and promote the Cr coating on Zircaloy-4 alloy autoclaved for 3024 h as the best corrosion resistance based on decrease in corrosion current density values simultaneously with the increase of the time spent in autoclave. A slow increase of Vickers micro hardness was observed as a function of the autoclaved period as well. The value reached for 3024 h being 219 Kgf/mm2 compared with 210 Kgf/mm2 value before autoclaving.
Highlights
The nuclear power generation industry has used Zr-based alloys, such as Zircaloy-2(Zr-2) and Zircalloy-4 (Zy-4) alloys, as fuel cladding materials in light water reactors owing to their neutron transparency and corrosion resistance [1,2]
Based on above recognized remarks, this paper presents the novelty of the testing corrosion behavior of Zy-4 coated with chromium by Thermionic Vacuum Arc (TVA) technique under CANDU reactor conditions
Based on the experimental data, we can present the following performances achieved after autoclaved for various periods of time of TVA Cr coatings of Zircaloy 4: a
Summary
The nuclear power generation industry has used Zr-based alloys, such as Zircaloy-2. (Zr-2) and Zircalloy-4 (Zy-4) alloys, as fuel cladding materials in light water reactors owing to their neutron transparency and corrosion resistance [1,2]. Various types of coatings have been proposed to improve the performance of current Zr-based alloy claddings, mainly regarding water side corrosion and high-temperature steam oxidation resistance, during normal and accident conditions. The recent relevant review devoted to protective coatings for accident tolerant Zr-based fuel claddings [20] introduced advanced latest results in the field of protective coatings for ATF cladding based on Zr alloys, presenting their behavior under normal and accident conditions in LWRs. The review attention has been focused on the protection and oxidation mechanisms of different coated cladding, including Cr coatings, as well as to the interdiffusion process between coatings and zirconium (these having less corrosion information available). It is shown that these coatings can be used as protective elements for existing fuel claddings, made of zirconium alloys, in light-water reactors of pressurized water reactor (PWR) and boiling water reactor (BWR) types
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