Modern measurement of local radial thermal conductivity of irradiated he-bonded U-10 wt. % Zr annular fuel: A cross-correlative study with electron probe microanalysis techniques

Abstract

The current study focuses on the thermal conductivity of irradiated U-10 wt. % Zr nuclear fuel, which is critical for the development of next-generation nuclear reactors, particularly sodium fast reactors. This work presents novel measurements of the local radial thermal conductivity of irradiated helium-bonded U-10 wt. % Zr annular fuel, utilizing thermoreflectance methods. In addition, cross-correlative approach was performed with microstructural and compositional measurements obtained from electron probe microanalysis (EPMA). The EPMA results indicate significant constituent redistribution within the fuel, with distinct inner and outer regions characterized by drastically different zirconium concentration profile than what is seen in irradiated Na-bonded U-based metallic fuel. The inner zone, enriched with zirconium, exhibited lower thermal conductivity due to the higher Zr concentration and further exacerbated by the presence of radiation induced defects, fission product nano-precipitates, phase segregation and nano-bubbles/voids. Conversely, the outer zone, depleted in zirconium, showed higher thermal conductivity values consistent with greater volume fractions of more conductive phases such as alpha uranium and U-Fe intermetallic phases (near the fuel-cladding interface). The study also reports on the thermal conductivity of irradiated HT-9 steel cladding, highlighting the impact of radiation-induced defects. This research provides critical insights into the thermal properties of irradiated U-10 wt. % Zr fuel, offering a first-of-a-kind data set which can be used in the development of predictive fuel performance models and the design of advanced reactor systems.