Mesoscale Thermophysical Property Characterization of Neutron Irradiated Tri-Structural Isotropic (TRISO) Fuel Particles

Measurement of the thermophysical properties of nuclear fuel systems is vital for advanced reactor research, development,
and deployment. As such, knowledge of these properties is necessary for modeling the performance of nuclear fuel. Here we
report the first post-irradiation characterization of changes in the thermal diffusivity and thermal conductivity of TRISO fuel.
The fuel in this study was irradiated at the Advanced Test Reactor at Idaho National Laboratory as part of the Advanced Gascooled Reactor (AGR) Fuel Development and Qualification program’s second test, AGR-2. Novel mesoscale measurements
of the thermal diffusivity of the TRISO fuel were performed using the INL-developed Thermal Conductivity Microscope, a
thermoreflectance-based microscope. Two irradiated compacts and three as-fabricated un-irradiated particles were measured
in this study and compared to literature data. In each fuel compact, roughly 3200 individual TRISO particle microspheres
(about 870 μm in diameter on average) were overcoated in a low-density graphitic matrix and pressed into a cylinder. TRISO
particles have a central (~430 μm) uranium oxycarbide (UCO) fuel kernel coated with a porous pyrolytic carbon (PyC) buffer
layer (~100 μm thick), an inner pyrolytic carbon (IPyC) layer (~40 μm), a silicon carbide (SiC) layer (~35 μm), and an outer
pyrolytic carbon (OPyC) layer (~40 μm). The fuel compacts were sectioned, then their cross-sectioned surfaces were
prepared by materiolographic grinding and polishing to a mirror finish. The thermal diffusivity was measured, and thermal
conductivity calculated using literature data, for each component of the TRISO particles in both radial and circumferential
directions (multiple TRSIO particles per compact measured to yield spatial resolution for the cross-section of the compact).
The two irradiated compacts had nearly identical burnup with an average of 11.35% Fissions per Initial Metal Atom, but
different Time-Averaged-Volume-Averaged (TAVA) irradiation temperatures of 1078ºC and 1216ºC. Significant differences
in the thermophysical properties and optical imagery of each of the components were observed between as-fabricated
particles and fuel particles post-irradiation. The irradiated UCO fuel particles exhibit increased porosity that contributes to
~57% reduction in thermal diffusivity, with some spatial variance radially through the particles. The buffer, inner pyrolytic
carbon, and outer pyrolytic carbon layers exhibited an increase in thermal diffusivity post-irradiation ranging from ~25% to
~50% compared to un-irradiated material. The pyrolytic carbons appear to exhibit densification that may likely contribute to
the aforementioned increase in thermal conductivity. The SiC layer exhibits a ~40% to ~60% decrease in thermal diffusivity,
with a higher thermal diffusivity appearing in the compact with the higher TAVA temperature. A degradation of the thermal
diffusivity, and hence thermal conductivity, in the components post-irradiation is expected due to the defect forming
mechanisms of irradiation, with defects acting as phonon-scattering sites to limit thermal wave propagation. It has been
shown in SiC and carbon/graphite that thermal diffusivity can recover with an increasing irradiation temperature due to
annealing that causes the recombination of defects. The system was analyzed for anisotropic effects, with some anisotropy in
the pyrolytic carbon layers, however these were within the uncertainty range of the measurements. Future work would
encompass more measurements including TCM, laser flash analysis and differential scanning calorimetry. Ultimately, the
currently reported data and future measurements will be used to support the qualification and performance optimization of
AGR fuel and associated reactor designs