A finite element approach to investigate the thermomechanical behavior of solid and annular type nuclear fuel rods

摘要: Abstract Reliable prediction of nuclear fuel rod behavior is prime concern for assessing proper design parameters as well operating conditions and sustainability aspects in a reactor. In this study, numerical analysis has been conducted to investigate the thermomechanical solid annular rods individually, considering pressurized water reactor (PWR). The transient heat conduction equation employed heat-generating uranium dioxide pellet along with surrounding zircaloy cladding solved numerically using finite element method. Loss coolant simulated by reducing transfer coefficient compared normal conditions. all cases considered, shown superior properties terms minimized surface flux, lower but relatively uniform radial temperature, thermal strain distribution.