Under current operating exposure times imposed by the five weight percent U-235 enrichment limit, the behavior of zirconium based fuel cladding alloys is reasonably understood on an empirical, and even somewhat phenomeno­logical, basis. This understanding is based on predictions of the oxide layer and hydride content of the cladding, both of which affect its ability to with­stand stresses due to normal and accident conditions. This does not mean that there are not unknown issues with the current zirconium alloy systems. Prediction of the cladding resistance to RIAs is not well understood; here large amounts of heat are almost instantaneously imposed on the fuel which causes rapid expansion of the pellets into the cladding resulting in clad­ding failure. As the exposure of the fuel increases and these rapid expan­sion effects become more pronounced and less well understood, the nuclear fuel manufacturers are left with the choice of understanding the mechanical behavior of highly exposed (and oxidized and hydrided) cladding to rapid stresses imposed by the fuel either on an empirical or phenomenological basis. The empirical basis used so far relies on the collection of immense amounts of data covering almost any potential event during fuel operation, a very time consuming and expensive approach. The phenomenological approach would be much better, but there is practically no such basis for understanding except at the very rudimentary level. Previous attempts to model and predict these effects have been unsuccessful due to the com­plexity of the interactions between the models of the various effects which lead to severe non-convergence problems. Major programs are currently underway funded by the U. S. Department of Energy (DOE) and industry to try to understand and model fuel behavior on a phenomenological basis using improved convergence algorithms (for instance the Consortium for Advanced Simulation of Light Water Reactors (CASL) and the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program).

Besides operating under extended burnup, the known operating limits of nuclear fuel cladding are being exceeded by high thermal duty condi­tions. This arises from the desire of nuclear utilities to get the most electrical generating capacity out of their current plants while still using a minimal amount of fuel. These higher thermal operating limits (in kilowatts per meter) increase the temperature of the cladding which, in turn, increases the build-up of crud on the surfaces, further increasing the cladding temper­ature and its oxidation rate and hydride content. Higher thermal rates also increase fretting issues between the fuel rods and their support structures causing unintended wear and fuel cladding failures. The added vibration is due not only to increased stress due to flow variations, but also to relaxation of the grid structures that support the fuel rods. In addition, there is a trend to move to slightly higher pH values by adding more lithium to the primary coolant. These higher lithium levels affect the corrosion rate of the fuel cladding though they also seem to inhibit stress corrosion cracking (SCC) of steam generator alloys in the primary circuit (EPRI, 2012).

Besides the more recent concern with RIA-type accidents and the ability to store used nuclear fuel for long times under both wet (spent fuel stor­age pool) and dry (spent fuel storage casks and perhaps internment in a final repository) conditions, there is the continuing concern of how fuel which has experienced high burnups will respond to such classic accident scenarios as large and small break loss of coolant accidents (LOCAs) due to loss in ductility and strength. Current knowledge of both oxidation layer thickness and hydride content is gained entirely through empirical testing. Development of new alloys is based entirely on an empirical approach with very little theoretical guidance.

Areas of research for zirconium (and any other metal) alloy claddings include the phenomenological understanding of: [22]

• The effect of oxide level and hydride content on the mechanical behav­ior of zirconium alloys.

• The effect of rapid stress levels on the mechanical behavior of zirconium alloys.

• Effects of long-term wet and dry storage as well as environmental condi­tions in any potential disposal site on the integrity of the zirconium alloy cladding.

• Effect of mechanical stresses induced by flow vibration on the fuel rod cladding as well as the grid support structures.

• Effect of flow rate on corrosion and erosion of both the fuel and the fuel structure.

• Modeling of these effects to allow performance prediction in extended burnup and in transient operating conditions.