Zircaloy-steam interactions during nuclear-fuel-rod heatup and melting

Research by the nuclear industry and regulating agencies has shown that the oxidation and melting of the Zircaloy cladding of nuclear fuel rods has a dominant influence upon the behavior of a reactor core during a severe accident like TMI-2. The work presented in this dissertation addresses three important aspects of that research. First, the influence of Zircaloy oxidation and melting upon overall nuclear fuel rod behavior is characterized through a critical review of experimental results and comparison of results from oxidation experiments with fuel-rod heating and melting experiments. Second, the fundamental mechanisms controlling the oxidation and melting of Zircaloy are defined through a critical review and discussion of oxidation experiments and supporting analytical work. Third, a finite elements model that treats the coupled heat conduction and oxygen diffusion within Zircaloy during heatup and melting, is presented and discussed. This model offers a significant extension of previous analytic capabilities with the incorporation of the Zr-ZrO2 binary phase diagram, algorithms to track the movement of a varying number of solid-solid or solid-liquid interfaces, and temperature - and concentration-dependent material properties