Neutronic analysis of CANDU fuel bundle with reprocessed fuel

Abstract

Fuel economy and novel fuel cycles for nuclear industry is a wide subject that required a massive effort along the decades. The reprocessing methods has shown a great potential to reach this goal to reduce the environmental issues related to burnt fuel disposal in nuclear resources utilization frame which is the primary subject. In second to reduce the mining and milling disadvantages which has minor effect if compared with the first. Besides the advantages reprocessing wastes should be considered. Thus, advanced nuclear fuel cycles has been proposed and one of the interest for CANDU technology is the communication between CANDU and LWR fuel cycles. In this concept, for short DUPIC/OREOX method is one of the most promising due cost and lower residual products and retain non-proliferation features. Since the spent PWR fuel has a higher fissile content than natural uranium, the DUPIC fuel results in a greater amount of fissile nuclides in fuel matrix. The objective of this paper is to state a roadmap to keep neutronics of the reprocessed fuel bundle as closer as possible to natural uranium fuel bundle to achieve a reliable scenario for further CANDU 6 reactor core fueling scheme. An evaluation of neutron absorber materials (depleted uranium, thorium, dysprosium and hafnium) was presented in sense of reactivity reduction due reprocessed fuel attributes. In the systems assessment the use of hafnium had shown to be suitable relating to system criticality. The SCALE 6.0 code was used for modeling and analysis.