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(2022)ThesisHigh burnup nuclear fuels have the potential to improve the economics of nuclear power, increase proliferation resistance and reduce waste generation. One technical challenge of increasing burnup is understanding the impact of the pellet-clad bond layer (PCBL) on heat transfer through the fuel. This thesis combines molecular dynamics simulations with experiments to investigate the thermal expansion, heat capacity and thermal conductivity of the PCBL. The PCBL was simulated as a (U,Zr)O2 solid solution. (U,Zr)O2 samples were synthesised using external gelation. It was found the using external gelation did not increase the solubility of ZrO2 relative to powder mixing methods and that to produce high density pellets (> 90 % theoretical density) of (U,Zr)O2 required hyperstoichiometry, with an O:M ratio > 2.10, and sintering temperatures exceeding 1500 C. Molecular dynamics (MD) simulations were performed on (U,Zr)O2 between 300 - 3000 K. The thermal expansion and heat capacity of all tested compositions was very similar up to 1400 K and consistent with the experimentally derived thermal expansion. Thermal conductivity was decreased for compositions containing ZrO2 at temperatures between 300--500~K, greater temperatures showed no discernible difference. The coefficient of thermal expansion, \(\alpha\), was measured experimentally for (U,Z)O2, with y = 0, 0.13 and 0.18, using in-situ XRD and was found to be 11-12 x10^-6 K^-1, for all three samples, at temperatures 300-600 K. No statistically significant difference in thermal expansion of the three samples was found, which matched results from the MD simulations. Phonon density of states (PDOS) measured using inelastic neutron scattering revealed that addition of ZrO2 caused a softening of phonons at 20~meV and a new phonon peak to appear at 23 meV. Additional simulations were performed on (U,Pu,Zr)O2. The addition of PuO2 causes a reduction in lattice parameter, albeit not as significant as with ZrO2 addition. PuO2 addition has little effect on the thermal expansion and heat capacity. PuO2 content caused an increase in thermal conductivity, although not enough to negate the impact of ZrO2. Further simulations on (U,Ce,Zr)O2 shows CeO2 to be a good surrogate for PuO2 with similar lattice parameter, heat capacity, thermal expansion and thermal conductivity.
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(2022)ThesisThe oxidation and H pick-up behaviour of Zr alloys used in the nuclear industry are dictated by alloying additions and coatings. This thesis presents computational findings on the kinetics of alloying element diffusion, formation and stability of secondary phases, and H trapping behaviour in these precipitates. In particular, it focuses on phases found in Fe- and Cr-containing Zr alloys, Nb-containing Zr alloys and Cr-coated Zr alloys. Density functional theory calculations were carried out to calculate the solubility of H in Zr(Fe,Cr)2 solid solutions with compositions representative of second phase particles found in Zircaloy-2 and Zircaloy-4. The results show that Zr(Fe,Cr)2 second phase particles are not strong traps for H. Combining these findings with experimental work carried out by colleagues, it is proposed that the interface between second phase particles and HCP-Zr matrix may act as possible trapping sites for H in Fe and Cr containing alloys. In Zr-Nb alloys, the solubility of H in β-Zr and β-Nb phases was calculated. It is found that the solubility of H increases with increasing Zr content in these β phases with H more soluble in β-Zr than in β-Nb. This explains the reported increase in terminal solid solubility of H in the presence of β-Zr in Zr-Nb alloys. A novel model is also presented, which predicts the solubility of H, to a remarkable level of accuracy, at any composition of (Zr,Nb) solid solution at a fraction of the computational cost required by current techniques. The transferability of the model to other systems and the inclusion of radiation-induced defects in the model are also discussed. Finally, the microstructural evolution of Cr-coated Zr alloy cladding is studied by calculating the vacancy-mediated diffusion of Zr and Nb solutes in BCC-Cr as well as Zr and Cr solutes in BCC-Nb. In BCC-Cr, both Nb and Zr are faster diffusers than Cr self-diffusion. Both Zr and Nb segregate towards vacancy sinks in BCC-Cr at normal reactor operating temperatures, but at elevated temperatures their flux is expected to be in opposite directions. In BCC-Nb, Cr is a slower diffuser than Nb self-diffusion, while Zr is faster; and both Zr and Cr are expected to decorate vacancy sinks in BCC-Nb at all relevant temperatures. The implications of these findings for Cr-coated Zr alloy cladding are discussed. The findings of this thesis can help the industry design alloys with lower H pick-up, allowing increased utilisation of nuclear fuel inside a reactor.