ANALYSIS OF DUPIC FUEL CYCLE USING THE MCNPX CODE.

Autores UPV
Año
CONGRESO ANALYSIS OF DUPIC FUEL CYCLE USING THE MCNPX CODE.

Abstract

The fissile content of the spent fuel from Pressurized Water Reactor (PWR) is about 1.5 wt%, which is higher than that of the fuel of CANDU reactor that uses natural uranium. Recent studies confirm the potential of using spent PWR fuel in the CANDU fuel cycle. In this case, an alternative to the once-through fuel cycle is the ¿Direct Use of spent PWR fuel In CANDU¿ (DUPIC), where the spent fuel from a PWR is packaged into a CANDU fuel bundle with only physical reprocessing (cut into pieces) but no chemical reprocessing. In this work, two techniques have been considered: AIROX (Atomics International Reduction Oxidation) and OREOX (Oxidation and REduction of OXide fuel). The goal is to evaluate the CANDU core behaviour based on the number of reprocessed fuels inserted. The keff and neutron flux at BOL (Begin of Life) and zero power conditions were analysed considering three core configurations: Standard CANDU core (natural uranium fuel), AIROX-CANDU core (natural uranium and AIROX fuel) and OEROX-CANDU core. The MCNPX code was used to model a CANDU-6 reactor core.