Nuclear reactors and power plants are such complicated and heterogeneous systems that they require complex physical and mathematical models to give a precise and clear description of their operation and accident characteristics. Till now, the development of nuclear thermalhydraulic analyses has evolved from initial coarse one-dimensional (1D) system codes such as RELAP5 [1], RETRAIN [2], CATHARE [3], and MARS [4] to finer revised component-scale codes COBRA [5], RELAP5-3D, VIPRE [6,7], and local-scale codes Fluent [8], CFX [9], and
Star-CCM [10]. There is no doubt that the above codes are still in state of the art; they have been widely used in the nuclear industry for operation simulation, safety analyses, and even severe accident phenomena analyses.