Pressurized water reactors (PWRs) are prevalent in nuclear power plants worldwide, contributing most of the electricity produced from nuclear energy. Modern PWRs allow subcooled flow boiling under their normal operating conditions to enhance the heat transfer rate. However, subcooled flow boiling might also suffer from critical heat flux (CHF) under accidental conditions, called departure from nucleate boiling, a major safety concern for the reactor’s design and operations. Thus, thermal-hydraulic analysis (THA) of subcooled flow boiling in PWRs is crucial in optimizing fuel assembly design to ensure efficient and safe operations. However, performing a THA in PWRs is a very complicated task because of the complexities involved in the operating conditions of the reactor, the physics of subcooled flow boiling, and subchannel geometry due to the presence of mixing vane grids (MVGs). In recent years, computational fluid dynamics (CFD) has emerged as a powerful tool for predicting subcooled flow boiling in various geometries, including PWR subchannels. The current study reviews CFD approaches for analyzing subcooled flow boiling in typical subchannels of a PWR under its prototypical operating condition or close to it. The impacts of various factors such as MVGs and their components (dimples, springs, and mixing vanes), cold walls, wall heat flux distributions, bowed rods, etc. on parameters such as lateral velocity, coolant pressure and temperature, void fraction, heat transfer coefficient, Nusselt number, wall heat flux partitions, CHF, etc. required to describe the behavior of coolant flow and heat transfer in subchannels, are presented. Finally, a summary of key conclusions and the scope for further research is presented.