Critical safety studies of a nuclear power plants are often associated with inadequate and improper cooling of the reactor core or the spent fuel rods. Coolant flow over the hot nuclear fuel rods often gets stalled during major accidents resulting in high temperature levels. These elevated temperature levels can potentially melt the fuel rod material and cause the release of radioactive gases. Research activities, both numerical and experimental in nature to explore these rare but potentially catastrophic possibilities have resulted in sophisticated numerical codes capable of simulating the various post-accident scenarios. These codes, although reasonably accurate and reliable have steep learning curves and are not often very user-friendly. A fast and accurate prediction of the critical temperature conditions using popular commercially available software packages is the subject of current study.
Results from this parametric study of temperature distribution over a partially cooled fuel rod carried out using ANSYS as the numerical analysis tool is reported. Nuclear fuel rods being inadequately cooled inside a stagnant pool of coolant water in an accident scenario resulting in disrupted coolant flow has been simulated. This situation can arise within the reactor (design-basis accidents) or in the waste-fuel storage (as faced in Fukushima). In these situations, the fuel rod is often left partially immersed in the coolant water resulting in immersed portion of the rod cooled by water and the exposed portion cooled by air leading to non-uniform and improper cooling of the system.
Realistic dimensions and materials as in commercial nuclear fuel rod have been used in the study. Taking advantage of the symmetry, an axisymmetric radial plane sliced longitudinally has been analyzed. Variations in the tangential direction have been neglected. The heat transfer problem uses homogeneous convective boundary conditions and assumes temperature dependent thermal conductivity. The parameters varied are the coolant level and the heat generation rate inside the fuel rod. A macro to automatically capture the transients in the temperatures was written in ANSYS (a finite element package). The governing energy equations were implicitly solved using finite volume scheme in MATLAB. ANSYS results are in close agreement with those obtained using MATLAB. The centerline temperature of the fuel rod shows a sharp rise below a certain coolant level.