To improve efficiency, safety, and reliability of nuclear reactors, structural materials for Gen-IV reactors are being designed and developed. Alloy 709, a 20Cr-25Ni austenitic stainless steel, has superior mechanical properties to be a preferred candidate material for Sodium Fast Reactor structural application. Creep tensile tests were performed at temperatures of 700 °C, 725 °C and 750 °C and range of stresses from 100 MPa to 250 MPa. The apparent stress exponent and activation energy were found to be 10.3±0.4 and 368.6±14.7 kJ/mol. Linear extrapolation method was used to rationalize the higher stress exponent and activation energy relative to the mechanism in power law creep yielding to a true stress exponent of 7.1 ± 0.3 and a true activation energy of 277 ± 12.8 kJ/mol which is close to the lattice diffusion of iron in Fe-20Cr-25Ni. Hence, the lattice diffusion controlled dislocation climb process is believed to be the rate controlling creep deformation mechanism in this range of stresses and temperatures. The appropriate constitutive equation was developed based on the results; however, microstructural evaluations are under investigation to confirm the rate controlling mechanism. In addition, creep tests at higher temperatures and lower stresses are being conducted to extend the stress and strain-rate ranges to observe possible transition in creep mechanism.
- Nuclear Engineering Division
Investigation on Creep Mechanisms of Alloy 709
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Alomari, AS, Kumar, N, & Murty, KL. "Investigation on Creep Mechanisms of Alloy 709." Proceedings of the ASME 2017 Nuclear Forum collocated with the ASME 2017 Power Conference Joint With ICOPE-17, the ASME 2017 11th International Conference on Energy Sustainability, and the ASME 2017 15th International Conference on Fuel Cell Science, Engineering and Technology. ASME 2017 Nuclear Forum. Charlotte, North Carolina, USA. June 26–30, 2017. V009T02A003. ASME. https://doi.org/10.1115/NUCLRF2017-3649
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