Use Of Lattice Code Dragon In Reactor Caluclations

The computer code Dragon is a free deterministic code developed by various organizations. It is a property of Ecole Polytechnique de Montreal. Dragon contains a collection of various models which can describe the neutron transport in a given geometry of a unit cell, reactor fuel assembly or in a reactor core. To obtain the final solution it is necessary to link together different modules at each step and any compromise at any level can lead to poor final results. For a nuclear engineer it is crucial to maintain the accuracy when reducing computational time. In the past the advanced self shielding models which were incorporated in the Dragon code Version4 were analysed. The conclusion obtained in that analysis was that the computational time of the burnup calculations was too long to be used for routine calculations. With the additional research and analysis presented in this paper satisfactory results were obtained that maintain the accuracy and reduce the computational time. In this paper these results will be presented. Results are compared to the reference results obtained by the Monte Carlo code SERPENT.