Nano-mechanical property assessment of a neutron-irradiated HT-9 steel cladding and a fuel-cladding chemical interaction region of a uranium–10 wt% zirconium nuclear fuel

Abstract

Nano-indentation was used in conjunction with electron microscopy to characterize the microstructure and mechanical properties of a neutron-irradiated HT-9 cladding adjoined to uranium–10 wt% zirconium, a fast reactor fuel. Electron microscopy revealed that the neutron-irradiated cladding can be classified into three localities: HT-9 edge, HT-9 + fission products, and the fuel-cladding chemical interaction (FCCI) region. The three localities possessed differing nano-hardness values and consequently resulted in differing calculated yield stresses. Following irradiation, the nano-hardness and yield stresses increased in the HT-9 edge. The HT-9 + fission products locality also increased in nano-hardness and yield stresses with average values similar to the HT-9 edge, but with larger standard deviations due to the diffusion of fission products along grain boundaries. The FCCI locality had the largest and most variable nano-hardness and yield stresses in comparison to the other regions. A yield stress comparison of the HT-9 edge and the literature of neutron-irradiated HT-9 is discussed. Graphical abstract: [Figure not available: see fulltext.].