| Abstract |
| The study was conducted to elucidate the dynamic strain aging behavior of Zr-1.5Nb-0.4Sn-0.2Fe alloy which was developed as an advanced nuclear fuel cladding material. Tensile tests were conducted in the range of 25℃ to 500℃ at the rate of 8.33×10^(-5)/sec and 1.67×10^(-2)/sec. The results showed that cladding exhibited the dynamic strain aging behavior above 200℃ which included the increase in the strength, work hardening exponent as well as the decrease in the ductility and strain rate sensitivity. The increase of heat treatment temperature from 470℃ to 510℃ shifted dynamic strain aging temperature to the higher range because initial dislocations causing dynamic strain aging were annihilated by the heat treatment. The analysis of the activation energy and diffusion model of solute atoms revealed that the mechanism of dynamic strain aging in Zr-1.5Nb-0.4Sn-0.2Fe alloy was due to the thermal migration of solute atoms such as oxygen with the help of the vacancy flow formed by substitutional atoms during plastic deformation. |
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| Key Words |
| Zirconium, Cladding, Dynamic strain aging, Solute atom, Diffusion |
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