Nuclear reactor fuel cell burnup calculations; an efficient method

A method of cell calculation for thermal neutrons has been developed for the simulation of a light water reactor fuel rod burnup. This method is based on the analytical solution of diffusion equation in the clad and moderator regions and numerical solution of integral transport equation in the fuel region. The accuracy of this technique has been verified by considering one energy group and two region fuel unit cell. The results of thermal disadvantage factor and neutron flux profile are obtained for various test problems. These results are then compared with results obtained from several other techniques of varying sophistication. Also included are the results of one energy group, three region unit cell. The present method provides satisfactory results, not only for the disadvantage factor, but also for the flux. This technique has been then implemented in the one dimensional burnup computer code MLASER, as well as the original LASER code. The new versions DTMLASER and DTLASER developed in this research have been applied to the sample problem supplied with the original LASER code. The new codes lead to a more efficient calculation resulting in saving of more than two-thirds of computational time for accuracy comparable to that of the previous codes