With efforts being made to prolong the burnup of nuclear fuels and increase
efficiency of pressurised water reactors (PWRs), there is a focus on extended
residence times of fuel. In addition, there are potentially significant cost benefits
through plant simplification if a soluble boron-free lithiated primary water chemistry can be demonstrated to be a viable route for PWR-based small modular reactor operation. However, the corrosion of the zirconium alloy clad material under lithiated conditions remains a concern as the mechanisms that underpin this have yet to be fully identified. Identifying the mechanism by which Li accelerates zirconium alloy corrosion will allow new alloying additions to be considered and new water chemistry regimes to be investigated, improving the efficiency and performance of future nuclear power reactors. Therefore, in this work the behavior of Li in ZrO2 by using density functional theory (DFT), to identify the most stable accommodation mechanisms for Li in ZrO2. A Brouwer diagram has been developed that predicts the nature of the defects present at given Li concentrations and partial pressures of oxygen. The solubility of Li in bulk ZrO2 is predicted to be low.