Heat Transfer in a Cross Flow Nuclear Reactor

The heat transfer between mutually perpendicular fuel and coolant channels of a nuclear reactor is studied for both isotropic and anisotropic conducting media. By means of an electrical conduction analog, a characteristic length of a unit cell of the reactor is experimentally determined for channels of rectangular and circular cross section. Empirical expressions for this characteristic length are given which closely fit the experimental data. A finite difference calculation is used to determine the temperature distribution within a unit cell.