Microstructural evolution in nuclear materials
Understanding the oxidation of zirconium alloys under irradiation
The pellets of uranium oxide fissile fuel in pressurized water reactors are contained within tubes of zirconium alloy - the fuel clad. Cladding materials present a particular opportunity for research with rapid impact, because they are amongst the few parts of a nuclear reactor that are replaced during its lifecycle. It is thus possible to make continual improvements to the performance of these materials as our understanding of them improves. The University of Manchester (UoM) has a particular strength in the materials science of zirconium-based fuel materials, led by the group of Prof. Michael Preuss.
Amongst other degradation processes, the zirconium alloy cladding undergoes oxidation in the high-temperature pressurized water in which it sits. A better understanding of the corrosion mechanism would allow for the design of safer, more efficient fuels. The pattern of corrosion involves repeated cycles of initially rapid then slower corrosion, before eventually a rapid breakaway phase, with linear kinetics, takes hold. Recent work at UoM suggests that this process is governed by a phase transition between two forms of the zirconium oxide. This phase transition is accompanied by a volume change that may promote cracking and the development of porosity and so permit more rapid transport of oxygen to the metal-oxide interface. Tin, present as an alloying element, appears to play a complicated role in the corrosion process, having both positive and negative effects on the corrosion resistance of the zirconium alloys.
Amongst other degradation processes, the zirconium alloy cladding undergoes oxidation in the high-temperature pressurized water in which it sits. A better understanding of the corrosion mechanism would allow for the design of safer, more efficient fuels. The pattern of corrosion involves repeated cycles of initially rapid then slower corrosion, before eventually a rapid breakaway phase, with linear kinetics, takes hold. Recent work at UoM suggests that this process is governed by a phase transition between two forms of the zirconium oxide. This phase transition is accompanied by a volume change that may promote cracking and the development of porosity and so permit more rapid transport of oxygen to the metal-oxide interface. Tin, present as an alloying element, appears to play a complicated role in the corrosion process, having both positive and negative effects on the corrosion resistance of the zirconium alloys.
Pellet-cladding interaction in fission reactor fuel
Zirconium alloys are key nuclear materials, used to form the cladding tubes that contain the uranium oxide fuel pellets in water-cooled fission reactors. Continual improvements to the reliability of the fuel clad will allow the design of better, more efficient fuels for the future. As we move towards higher burn-up in harsher environments, a proper mechanistic undestanding of fuel clad failure mechanisms will be critical to enabling these improvements .
The process of nuclear fission of the uranium oxide fuel produces iodine. This iodine is released from the fuel pellets, particularly during shut-downs for refuelling. The iodine then interacts with the zirconium alloy cladding reducing its strength. When the reactor is powered back up, stresses in the fuel clad arise and cracks can form. This iodine-induced stress corrosion cracking (SCC) is a key mode of failure of fuel rods in what is called the pellet-cladding interaction (PCI). We do not yet have a detailed mechanistic understanding of PCI failure.
The process of nuclear fission of the uranium oxide fuel produces iodine. This iodine is released from the fuel pellets, particularly during shut-downs for refuelling. The iodine then interacts with the zirconium alloy cladding reducing its strength. When the reactor is powered back up, stresses in the fuel clad arise and cracks can form. This iodine-induced stress corrosion cracking (SCC) is a key mode of failure of fuel rods in what is called the pellet-cladding interaction (PCI). We do not yet have a detailed mechanistic understanding of PCI failure.