Graphite-boron composites for enhanced thermal conductivity and irradiation shielding for use in spent nuclear fuel dry casks

This is the authors' abstract. We don't add key points for this paper.

Graphite-boron composite pellets with 1:1 weight ratios were produced by uniaxial compaction with a phenolic resin binder, followed by heat treatments. Both B 4 C and BN were utilized for manufacturing the composites. These composites exhibited enhanced thermal conductivities owing to the graphite content and anisotropic properties owing to the uniaxial compaction method. At 300 ℃, the radial direction demonstrated a thermal conductivity increase of up to three times compared to that at the axial direction and against SUS304, the material conventionally utilized for canisters in spent nuclear fuel dry casks. Geant4, Monte Carlo simulation code, confirmed that more than 97% of thermal neutrons were inhibited by just 2 mm thick composites owing to their boron content, while SUS304 hindered only 20% of neutrons under identical conditions. This study highlights the promising thermal conductivity and radiation shielding properties of graphite-boron composites, and future research will focus on improving their density and mechanical properties.