Nuclear reactor fuel burnup and transport calculations

A computer code TWO LASER has been developed for the neutronic calculation of square lattice cells in nuclear power reactors. This computer code, which uses new methods for solutions of the integral transport equation and bunup equations, has been used to assess the accuracy and speed of the LASER code, which uses a one-dimensional approximation to the actual two-dimensional system. Calculations have been performed on a sample problem for a burnup of 10.6 MWD/kg fissile. Results from these calculations show that the one-dimensional approximation used by LASER is good for the sample problem. However, the method used by LASER for the solution of burnup equations is not efficient. A modified version of the LASER code has also been developed in this research. The modified LASER, which used the one-dimensional approximation of LASER and the new method for the solution of burnup equations, gives good results compared to the two-dimensinal code. However, the modified LASER is more efficient as far as computional time is concerned. The new computer code TWO LASER can be used to profice data for benchmarking, and the modified LASER can be used for the replacement of the original LASER