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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 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.

Carbon-14 (radiocarbon,14C) is a long-lived radionuclide (5730 yr) of interest regarding the safety for the management of intermediate level wastes (ILW). The present study gives an overview of the release of14C from irradiated Zircaloy cladding in alkaline media.14C is found either in the alloy part of Zircaloy cladding due to the neutron activation of14N impurities by14N(n,p)14C reaction, or in the oxide layer (ZrO2) formed at the metal surface by the neutron activation of17O from UO2or (U-Pu)O2fuel and water from the primary circuit in the reactor by17O(n,α)14C reaction. Various irradiated and unirradiated Zircaloys have been studied. The total14C inventory has been determined both experimentally and by calculations. The results seem to be in good agreement. Leaching experiments were conducted in alkaline media for several time durations.14C was mainly released as carboxylic acids. Further, corrosion measurements were performed by using both hydrogen measurements and electrochemical measurements. The corrosion rate (CR) ranges from a few nm/yr to 100 nm/yr depending on the surface conditions and the method used for measurement. From a safety assessment point of view, the instant release fraction (IRF) was determined on irradiated Zircaloy-2. The results showed that the14C inventory in the oxide was significantly below the 20% commonly used in safety case assessments.