The relationship between mechanical properties and microstructure of HT9 steel

Next-generation advanced reactor system designs present new challenges for material design and selection. The structural materials selected for these reactors will need to withstand higher operation temperatures, more neutron irradiation (greater flux), and corrosion from liquid metal coolants. HT9, a ferritic/martensitic alloy, was used in previous fast research reactors (such as EBR-II and FFTF). However, it needs validation before it can be selected for use in the more extreme environments of Gen–IV reactors. This dissertation presents a series of in-situ high-energy synchrotron X-ray diffraction (XRD) tensile tests conducted on alloy HT9. The loading behavior of HT9 was examined using diffraction line profile analysis methods. Analysis of the shift in diffraction peak position during deformation of the specimens allowed for the determination of elastic lattice strains in the two primary constituent phases of the material: the ferritic/martensitic matrix and the Cr23C6 carbide particles. With the initiation of plastic deformation, the samples exhibited a clear load transfer from the matrix phase to the carbide particulate. The evolution of the dislocation density in the material as a result of deformation was characterized by peak broadening analysis. Small scale tensile samples of HT9 were strained at room temperature and at test temperatures from 300 C to 500 C. The Cr23C6 carbide phase in the material is shown to accommodate a significant portion of the load after the ferrite phase yields, despite making up only 3% by volume. A set of irradiation-damaged HT9 samples harvested from a duct irradiated in the FFTF were also examined using in-situ XRD. This unique set of specimens extracted from the ACO–3 duct represents a first set of samples to be irradiated under realistic time-variant conditions, including cyclic temperature and fluence variations