Abstract : If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxyde fuel in present-day light water reactors.
This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions.
Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors.