Mechanical properties and thermal conductivity of (U,Zr)SiO4

Abstract

During the core meltdown at the Fukushima Daiichi Nuclear Power Plant (1F), molten core materials penetrated the reactor pressure vessel (RPV), and they may have reacted with the concrete at the bottom of the primary containment vessel (PCV). Through this molten core concrete interaction (MCCI), uranium could potentially dissolve into ZrSiO4 and form a (U,Zr)SiO4 solid solution. For fuel debris retrieval to proceed smoothly, detailed information on the physical properties and critically characteristic of all possible fuel debris, including MCCI products, is necessary. However, no studies have reported on the physical properties of uranium in ZrSiO4 because it is difficult to synthesize. In this study, we fabricated dense (U,Zr)SiO4 samples through the spark plasma sintering (SPS) of SiO2 and (U,Zr)O2 and analyzed the solubility of uranium in ZrSiO4. The solubility limit of uranium in ZrSiO4 was found to be within 1.6 – 2.6 at%. The mechanical properties of the (U,Zr)SiO4 solid solution were calculated using the samples’ sound velocities. To understand the effect zirconium substitution by uranium has on heat transport, the thermal conductivity was calculated using the thermal diffusivity data measured with laser flash. This is the first study on the thermophysical properties of (U,Zr)SiO4, and it provides us with a better understanding of the properties of MCCI products.