Abstract
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.
Highlights
IntroductionAt the end of last century in the environment, which needed more safe and efficient energy sources, such a concept was enforced
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
Materials chosen for reactors must have acceptable dimensional stability under radiation, be resistant to irradiation creep under constant radioactive stress, possess acceptable ductility, toughness, creep rupture strength, and chemical compatibility against stress corrosion cracking (SCC), and irradiation assisted SCC with supercritical water [7,8]
Summary
At the end of last century in the environment, which needed more safe and efficient energy sources, such a concept was enforced In this context, at the beginning of our century, the Generation IV International Forum was initiated to develop advanced nuclear reactor designs to improve sustainability with safety, reliability, and economical energy, due to better resistance and physical protection [1,2]. 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 a supercritical water-cooled nuclear reactor (SCWR). The available data about changing the mechanical properties of austenitic steels after irradiation present how such alloys exhibit hardening up to approximately
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