抄録
This paper performed the transient core analysis of a small molten salt reactor in the case of blockage accident. The emphasis is that the numerical model developed in this paper takes into account the interaction between fission reaction and fuel salt flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis in the case of blockade accident are as follows ; (1) neutron multiplication factor hardly changes, (2) the outlet temperature of fuel salt decreases 10 K, and (3) fuel salt and graphite temperatures largely increase at the blockage point, but lower than fuel salt boiling temperature and the molten temperature of the reactor vessel.
