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 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.

In the nuclear industry, coated cladding is a topical problem and it is chosen as the near-term and most promising ATF (Accident-Tolerant Fuel) cladding concept. The main objective of this concept is to enhance the accident tolerance of nuclear power plants and accordingly, the performance of cladding is expected to be improved. This work assesses the corrosion performance of a Zircalloy-4 alloy coated with a thin chromium coating by MS (magnetron sputtering), tested under a CANDU (CANada Deuterium Uranium) reactor primary circuit simulated condition (LiOH solution, 10 MPa, 310 °C, pH = 10.5). The anticorrosive performance is evaluated by a gravimetric analysis, a metallographic analysis, X-ray diffraction, electronic microscopy, and electrochemical methods. A four times less gain mass was noticed compared to uncoated Zircaloy-4, indicating a smaller corrosion rate. The SEM micrographs illustrate that the coatings are still adherent, and they are keeping the initial morphological characteristics during the autoclaving process. A SEM cross-section analysis shows values of the thickness of the coatings between 0.8 and 1.46 µm. By XRD, the presence of Cr₂O₃ oxide is identified. Electrochemical testing confirms good stability and good corrosion performance of Cr coating over time under autoclave conditions.

The nuclear industry is focusing some efforts on increasing the operational safety of current nuclear reactors and improving the safety of future types of reactors. In this context, the paper is focused on testing and evaluating the corrosion behavior of a thin chromium coating, deposited by Electron Beam Physical Vapor Deposition on Zy-4. After autoclaving under primary circuit conditions, the Cr-coated Zy-4 samples were characterized by gravimetric analysis, optical microscopy, SEM with EDX, and XRD. The investigation of the corrosion behavior was carried out by applying three electrochemical methods: potentiodynamic measurements, EIS, and OCP variation. A plateau appears on the weight gain evolution, and the oxidation kinetics generate a cubic oxidation law, both of which indicate a stabilization of the corrosion. By optical microscopy, it was observed a relatively uniform distribution of hydrides along the samples, in the horizontal direction. By SEM investigations it was observed that after the autoclaving period, the coatings with thickness from 2 to 3 µm are still adherent and maintain integrity. The XRD diffractograms showed a high degree of crystallinity with the intensity of chromium peaks higher than the intensity of zirconium peaks. Electrochemical results indicate better corrosion behavior after 3024 h of autoclaving.