The Molten Salt Reactor (MSR), proposed along with other five innovative concepts of fission nuclear reactor by the Generation IV International Forum (GIF-IV), represents a challenging task from the modelling perspective because of the strong coupling between neutronics and thermo-hydrodynamics due to liquid fuel circulation in the primary loop. In this paper COMSOL Multiphysics® is adopted to investigate the MSR neutronics, focusing on the steady-state core average conditions of the Molten Salt Breeder Reactor (MSBR) developed at Oak Ridge National Laboratory (ORNL). The results achieved by COMSOL, adopting a two energy group diffusion model and using group constants calculated by means of the deterministic code SCALE5.1, are compared with those achieved by the stochastic code MCNP for validation purpose. In particular, neutron flux profiles and integral quantities, like the effective multiplication factor and homogenized cross sections, are evaluated and discussed. The model implemented in COMSOL is then used to study the effect of the fuel velocity on the neutronic behaviour of the analysed MSBR core channel.
A Preliminary Approach to the Neutronics of the Molten Salt Reactor by means of COMSOL Multiphysics
MEMOLI, VITO;CAMMI, ANTONIO;DI MARCELLO, VALENTINO;LUZZI, LELIO
2009-01-01
Abstract
The Molten Salt Reactor (MSR), proposed along with other five innovative concepts of fission nuclear reactor by the Generation IV International Forum (GIF-IV), represents a challenging task from the modelling perspective because of the strong coupling between neutronics and thermo-hydrodynamics due to liquid fuel circulation in the primary loop. In this paper COMSOL Multiphysics® is adopted to investigate the MSR neutronics, focusing on the steady-state core average conditions of the Molten Salt Breeder Reactor (MSBR) developed at Oak Ridge National Laboratory (ORNL). The results achieved by COMSOL, adopting a two energy group diffusion model and using group constants calculated by means of the deterministic code SCALE5.1, are compared with those achieved by the stochastic code MCNP for validation purpose. In particular, neutron flux profiles and integral quantities, like the effective multiplication factor and homogenized cross sections, are evaluated and discussed. The model implemented in COMSOL is then used to study the effect of the fuel velocity on the neutronic behaviour of the analysed MSBR core channel.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
