Tensile and Corrosion Behavior of Zr705 for Nuclear Hydrogen Generation
Zr705 has been evaluated for its metallurgical and corrosion properties for application in nuclear hydrogen generation using a thermochemical process that involves the decomposition of hydroiodic acid at temperatures up to 400 °C. The results indicate that the tensile strength was gradually reduced with increasing temperature. However, the failure strain was enhanced up to a critical temperature (200 °C) followed by its reduction beyond it, possibly due to the dynamic strain aging effect. As to the cracking susceptibility in an acidic solution, no failure was observed under constant-load and self-loaded conditions. However, enhanced ductility was observed in slow-strain-rate (SSR) testing in an identical environment at elevated temperatures. The application of external potential during SSR testing enhanced the cracking susceptibility. The critical potentials obtained in the electrochemical polarization study became more active with increasing temperature. Ductile failures, characterized by dimples, were noted in all failed specimens.
Applied potential; Fractography; Stress corrosion; Stress corrosion cracking; Zirconium alloys – Fatigue; Zirconium alloys – Fracture; Zr705
Materials Science and Engineering | Mechanical Engineering | Mechanics of Materials | Metallurgy | Nuclear Engineering
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Roy, A. K.,
Kaiparambil, A. V.
Tensile and Corrosion Behavior of Zr705 for Nuclear Hydrogen Generation.
Materials Science and Engineering A, 427(1-2),