Ultrafine-grained oxide-dispersion-strengthened 9Cr steel with exceptional strength and thermal stability

As important structural materials for sodium-cooled fast reactors and fusion reactors, oxide-dispersionstrengthened (ODS) steels of the ferritic/martensitic (F/M) type containing 8-12 wt% Cr have been extensively studied. The F/M matrixes in these ODS steels, however, are often microcrystalline (between similar to 1 and 10 mu m). Decreasing the grain size to an ultrafine scale (0.1 - 1 mu m) should increase both the strength and the radiation tolerance. Here, by means of mechanical alloying and high-pressure/high-temperature consolidation techniques, we prepared novel ultrafine-grained (UFG) hafnium-doped 9Cr ODS steels. The UFG 9Cr ODS steels are composed of matrix(es) with an average grain size of similar to 0.17 mu m and high-density (5 x 10(22) m(-3)) hafniumenriched oxide nanoprecipitates with an average particle size of similar to 6 nm. No significant grain growth is observed in UFG 9Cr ODS steel annealed at 1000 degrees C. In contrast, the grain of previous UFG T91 alloys is stable below 500 degrees C. The extremely high thermal stability is caused by the combined kinetic and thermodynamic stabilization effects, i.e., the grain growth is kinetically hindered by the high-density oxide nanoprecipitates and thermodynamically hindered by the lowered grain boundary energy resulting from the segregation of elemental Mo at grain boundaries. The yield strength of UFG 9Cr ODS steel is 2449 MPa, approx. twice that of previously reported microcrystalline 9Cr ODS steels and UFG T91 alloys. The high yield strength is mainly caused by dislocation and grain boundary strengthening effects. The strategies found in this study should help design UFG F/M ODS steels with enhanced structural stability and mechanical property for nuclear reactor applications.

Automated summary

Novel ultrafine-grained hafnium-doped 9Cr ODS steels with high thermal stability and yield strength were successfully prepared, which could contribute to the design of more stable and mechanically robust materials for nuclear reactors.