In this work, the PSI FAST code system is extended for the modeling and simulation of Molten Salt Reactor (MSR) dynamics. The thermal-hydraulic code TRACE has been provided with a module for 1-D Delayed Neutron Precursor (DNP) balance and decay heat modeling in fluid fuel, and with built-in MSR materials. The reactor power is determined by means of a Point-Kinetics approach that makes use of power-weighted values of temperature and DNP distributions in the core. To validate the module, models for the comparison with experiments performed in the Molten Salt Reactor Experiment (MSRE) have been developed. The MSRE was a graphite moderator reactor built and operated in the sixties at Oak Ridge National Laboratory (ORNL). The neutronic characteristics of the reactor have been determined by means of the Monte Carlo code SERPENT. Two models of the MSRE plant with different detailing have been set up in order to determine the importance of the plant components. In particular, different descriptions of the external cooling loop have been tested and compared. Some significant transients have been considered for the module assessment, by comparison with available experiment data from ORNL, both in time and frequency domain.