Abstract
Understanding creep properties and microstructural evolution for candidate materials of the next-generation nuclear reactors is essential for design and safety considerations. In this work, creep tests were carried out at temperatures ranging from 973 to 1073 K and stresses 40 to 275 MPa followed by microstructural examinations of a Fe-20Cr-25Ni (mass pct) austenitic stainless steel (Alloy 709), a candidate structural material for the Sodium-cooled Fast Reactors. The apparent stress exponent and activation energy were found to be 6.8 ± 0.4 and 421 ± 38 kJ/mole, respectively. The higher activation energy relative to that of lattice self-diffusion together with the observation of dislocation-precipitate interactions in the crept specimens was rationalized based on the concept of threshold stress. The threshold stresses were estimated using a linear extrapolation method and found to decrease with increased temperature. By invoking the concept of threshold stresses, the true stress exponent and activation energy were found to be 4.9 ± 0.2 and 299 ± 15 kJ/mole, respectively. Together with the observation of subgrain boundary formation, the rate-controlling mechanism in the Alloy 709 was conclusively determined to be the high-temperature dislocation climb. Three types of precipitates were identified in the crept samples: Nb(C, N),
Z
-phases of sizes between 20 and 200 nm within the matrix and M
23
C
6
with sizes between 200 and 700 nm within the matrix and on grain boundaries. Further, the analysis of creep rupture data at high stresses indicated that the Alloy 709 obeyed Monkman–Grant and modified Monkman–Grant relationships with creep damage tolerance factor of ~ 5. Using the Larson–Miller parameter, it was concluded that the Alloy 709 exhibited superior creep strengths relative to the other advanced austenitic steels.