抄録
A new approach for the numerical simulation of the Molten Salt Reactor (MSR) is described in this paper. The MSR is a thermal neutron reactor with graphite moderation and employs fissile salt as fuel material. Within the MSR, the fission reaction, the fuel salt flow and the heat transfer are inseparably linked each other due to fission fuel circulation in the reactor. A numerical model developed here consists of two-group diffusion equations for fast and thermal neutron fluxes, balance equations with convective terms for six-group delayed neutron precursors, an energy equation for the heat convection and conduction of fuel salt and an energy equation for the heat conduction of graphite moderator. These equations were iteratively solved by the SOR algorithm. The verification of the model of the fission reaction was performed using the Standard Reactor Analysis Code (SRAC). The results showed that the model has comparable accuracy with the SRAC. Moreover, the interactions of the fission reaction, the fuel salt flow and the heat transfer were analyzed. As a result, the fuel salt flow affected the distributions of the precursors significantly, but those of neutron fluxes slightly. And both the precursors and the neutron fluxes decreased in the high temperature region of the reactor. In contrast, they increased in the low temperature region.
