| The main objectives in this research are to build, to develop and to write a computer program that will be used for modeling, for simulating and for safety analysis of the nuclear fuel channel in the CANDU reactor core following hypothetical accidents. To assess the ability to simulate the CANDU core, validation and improvement of this program are conducted with large number of simulations. Important comparisons between the fluid flow model predictions and the experimental data obtained at the IGN thermal loop were carried out in up-flow in heated test sections. The results show the good agreement between the predictions and the data. In addition, a complete simulation of a horizontal fuel channel with a non uniform heat flux (co-sinus) as the same nominal operation conditions of CANDU-6 nuclear reactor are also compared with the analytical solution. The simulations are in perfect agreement with the analytical solutions. (Abstract shortened by UMI.)... |
