The meltdown of a fuel rod is a severe accident resulting from the overheating of the reactor core. In the present study, a numerical investigation of this process, focusing on the loss of coolant has been conducted. The objective of this study is to conduct a numerical simulation of transient heat conduction and melting at various points within a typical pressurized water reactor fuel rod. In this analysis, heat conduction in the radial direction of a fuel rod, including the UO2 fuel pellet, the gap, and the zircaloy cladding, is investigated. The FRAPCON steady-state code is employed to calculate the operational parameters of the fuel rod. The calculated parameters, such as coolant and fuel temperatures, fission gas fraction, gap heat transfer coefficient, and burnup, are utilized to evaluate and compare the melting phenomena at different time intervals.

In the investigation of the phase change in various parts of the pellet and fuel rod, the explicit finite difference (FD) method is utilized with enthalpy instead of temperature-dependent equations. Finally, the temperature history, phase change, and melting map at different points along the radial and axial directions of the fuel rod during coolant loss and heat transfer coefficient reduction are evaluated based on various operating parameters of the core. To enhance the quality of the results, an uncertainty analysis of effective parameters is conducted.

According to this analysis, the heat transfer coefficient of the coolant under accident conditions (0.2 ± 5% kWm−2K−1) and the thermal conductivity of the fuel have the most significant impact on the temperature history and melting process. Highlights include the following:

1. The meltdown of a nuclear fuel rod is analyzed under a loss-of-coolant accident.

2. The enthalpy formula is discretized by the explicit FD numerical method.

3. Effective parameters in melting, such as coolant temperature, burnup, and gap heat transfer coefficient, are obtained by FRAPCON.

4. The temperature history, phase changes, and melting map of various radial points within the fuel pellet and cladding along the axial direction of the fuel rod are determined.