Avaliação neutrônica do uso de combustível reprocessado em um reator CANDU

Abstract

The CANadian Deuterium Uranium or CANDU (specifically CANDU 6) is a reactor developed by Atomic Energy Canada Limited - AECL that uses heavy water as moderator and coolant and natural uranium as fuel. The Korea Atomic Energy Research Institute (KAERI), in collaboration with AECL and US Department, work in development of advanced nuclear fuel cycles in sense of use spent fuel from Light Water Reactors (LWR) in CANDU, namely Direct Use of spent PWR fuel In CANDU (DUPIC) or DUPIC cycle. The spent fuel of a Pressurized Water Reactor (PWR) contains about 1.5wt% of fissile material depending on the burnup in the PWR. Due this feature there are possibilities in use of this material in the CANDU reactors. In this way, among the possible methods of reprocessing, the Oxidation and REduction of OXide Fuel (OREOX) that is a dry redox process that takes out only gaseous fission products, is the most promising technique due the reduction of costs and processes. This work evaluates the DUPIC cycle for a typical CANDU 6 reactor where the spent fuel from a PWR was reprocessed by the OREOX method. The use of reprocessed fuel causes an increase in the reactivity value in reactor core, so the use of a neutron absorber material is necessary to decrease the criticality. In this work the use of depleted uranium as absorber was proposed in order to compensate the increase in reactivity and to obtain a core configuration that presents neutronic behavior similar to CANDU 6 without any modifications (reference). The study was divided into two stages, the first one evaluated the initial criticality in steady state for several core configurations. The final phase the criticality and evolution of the composition during the burning. The best core configuration obtained from the analisys was 81.5wt% natural uranium, 6.9wt% depleted uranium and 11.3wt% reprocessed fuel with content around of 7kg of fissile material per metric ton of heavy metal higher than the CANDU 6 reference at the end of the burnup. The SCALE 6.0 (Standalone Computer Analyzes for Licensing Evaluation Modular Code) was used to analyse the initial criticalities and burnup in different combinations of natural uranium, depleted uranium and reprocessed fuel-core setups.