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.

The present paper is a narrative review focused on a few important aspects and moments of trends surrounding materials and methods in sustainable nuclear energy, as an expression of applied chemistry support for more efficiency and safety. In such context, the paper is focused firstly on increasing alloy performance by modifying compositions, and elaborating and testing novel coatings on Zr alloys and stainless steel. For future generation reactor systems, the paper proposes high entropy alloys presenting their composition selection and irradiation damage. Nowadays, when great uncertainties and complex social, environmental, and political factors influence energy type selection, any challenge in this field is based on the concept of increased security and materials performance leading to more investigations into applied science.

The aim of this work was to study the corrosion behavior of a Fe-Cr-Ni alloy (310 H stainless steel) in water at a supercritical temperature of 550 °C and a pressure of 250 atm for up to 2160 h. At supercritical temperature, water is a highly aggressive environment, and the corrosion of structural materials used in a supercritical water-cooled nuclear reactor (SCWR) is a critical problem. Selecting proper candidate materials is one key issue for the development of SCWRs. After exposure to deaerated supercritical water, the oxides formed on the 310 H SS surface were characterized using a gravimetric analysis, a metallographic analysis, and electrochemical methods. Gravimetric analysis showed that, due to oxidation, all the tested samples gained weight, and oxidation of 310H stainless steel at 550 °C follows parabolic rate, indicating that it is driven by a diffusion process. The data obtained by microscopic metallography concord with those obtained by gravimetric analysis and show that the oxides layer has a growing tendency in time. At the same time, the results obtained by electrochemical impedance spectroscopy (EIS) measurements indicate the best corrosion resistance of Cr, and (Fe, Mn) Cr₂O₄ oxides developed on the samples surface after 2160 h of oxidation. Based on the results obtained, a strong correlation between gravimetric analysis, metallographic analysis, and electrochemical methods was found.

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.

Today, countries all over the world, faced with a global energy crisis and the effects of climate change, are looking for alternatives to fossil fuels