Design and performance assessment of the molten salt fast reactor

In this thesis, COMSOL v 5.2 Multi physics software was utilized to perform coupled neutronics and thermal-hydraulics simulations of a molten salt fast reactor, and the Scale v 6.1 code package was utilized to generate the homogenized cross-section data library. The library's 238 cross-section groups were categorized into nine groups for the simulations in this study. The results of the COMSOL model under no fuel flow conditions were verified using the Scale v 6.1 code result, and the results of the neutronics and thermal-hydraulics simulations were compared to the results of previously published studies. The results indicated that the COMSOL model that includes the cross-section library generated by the Scale v 6.1 code package is suitable for the steady-state analysis and design assessment of molten salt fast reactors. Subsequently, this model was utilized to investigate the neutronics and thermal-hydraulics behaviours of the reactor. Multiple designs were simulated and analyzed in this model, and the results indicated that even if the wall of the core is curved, hotspots occur in the upper and lower portions of the core's centre near the reflectors. A new design was proposed that utilizes a flow rate distribution system, and the simulation results of this design showed that the maximum temperature in the core was approximately 1032 K and no hotspots occurred but the maximum velocity is nearly equal to 7 m/s. The new design of the molten salt fast reactor was modified again to reduce the maximum magnitude velocity. The results indicated that the new design reduces the magnitude velocity with acceptable hot spot temperature where the maximum magnitude velocity is nearly equal to 4.59 m/s and the hot spot temperature is less than 9 C above the average core outlet temperature. The performed analyses are used to adjusting the amount of fissile material started for k_eff=1. Neutron flux and delayed neutron distribution are studied where the position of maximum neutron flux is at the centre of the core and the position of maximum thermal neutron flux is the fertile blanket. Different transients will be performed such as unprotected transient overpower,unprotected loss of flow, unprotected loss of heat sink, and unprotected over-cooling are intensively investigated and discussed with the new core geometry. It is clear that due to negative reactivity feedback, an MSFR reactor has a high safety potential. The best reprocessing method was determined and evaluate the change of heavy nuclides with time. The value of the Breeding Ratio at the start is determined. The average net ²³³U production rate and Doubling Times are determined