Análise neutrônica dos reatores CANDU e ACR

Abstract

Nuclear reactors are widely used to generate electricity in many countries. Argentina, China, Canada, South Korea, India, Pakistan and Romania use heavy water (D2O) moderated and cooled reactors, using natural uranium as fuel. The CANDU (Canadian Deuterium Uranium) is an example of this type of reactor. Currently, a variation of CANDU, the ACR (Advanced CANDU Reactor), is in the design phase in countries such as Canada and South Korea. While CANDU uses D2O as refrigerant and moderator, ACR employs H2O as refrigerant and D2O as moderator. This amendment aims to reduce the volume of D2O, which is a more expensive resource from an economic point of view. As a consequence, the ACR must be fed with slightly enriched uranium. In this work, the insertion of reprocessed fuel in the ACR that has a percentage of fissile material greater than natural uranium is evaluated. The objective is to evaluate the neutron behavior of the CANDU-6 and ACR-1000 reactors, using the MCNPX 2.6.0 code (Monte Carlo N-Particle Extended – version 2.6.0). The ACR-1000 was evaluated considering different fractions of reprocessed fuel in the reactor core. This methodology aims to obtain keff values ​​closer to those of traditional CANDU. The results show positive characteristics of the ACR-1000 compared to its predecessor, as well as the use of reprocessed fuel.