Neutronics methods for the multiphysics analysis of nuclear fission systems

The accurate representation of the overall behaviour of advanced nuclear fission systems, including both nuclear reactors and accelerator-driven subcritical systems, typically requires to account for the interactions among the various physical phenomena that occur in the complex system comprised by a nuclear reactor. Correspondingly, the numerical codes that are intended for application to the design and the analyses of such systems are required to couple individual physics models, perhaps most notably those for neutronics and thermal-hydraulics, as these two phenomena are of great importance for safety considerations. The doctoral research activity involves the development of physical models and mathematical methods for spatial-temporal analyses of neutronics that are appropriate for use in the context of multiphysics analyses of nuclear reactor dynamics. The activity is carried out through the management, the development and the application of a multiphysics reactor analysis code, the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code, developed at Politecnico di Torino. First, for questions related to the practicality of the execution of spatial kinetics analyses, the quasi-static method is formulated and methods of solution of the resulting systems of equations are developed and characterised. Next, as the accuracy and the efficiency of the quasi-static approach depend on the use of proper time steps, an automatic, adaptive time step selection methodology for the quasi-static method is developed and studied. Additionally, as the effects of photon energy deposition lead to non-negligible effects on the spatial-temporal behaviour of the power in a nuclear reactor, a methodology for coupled neutron-photon dynamics appropriate for use in spatial kinetics analyses is developed and assessed. Finally, with these models and methods implemented in the FRENETIC code, the first coupled neutronic/thermal-hydraulic validation of the FRENETIC code is performed against experimental data from a formerly operating sodium-cooled fast reactor