DESIGN OF MOLTEN SALT REACTOR 100MWE TO BE APPLIED IN INDONESIA USING CITATION AND MCNP

• Main Author: Wulandari, Cici
• Format: Dissertations
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/70302
• Institution: Institut Teknologi Bandung

In this research, the neutronic design of the Molten Salt Reactor (MSR) has been performed with an operating power electric of 100 MWe. The selection of low power, 100 MWe, aims to make this reactor be used in areas with low population, for example, outside Java Island such as West Kalimantan or small islands that have a problem with the disconnected electricity grid. MSR is one of the generation IV reactor concepts which is fueled by molten salt. The fuel schemes used in this study include Thorium-Uranium enriched (Th-U), Thorium-Plutonium (Th-Pu), and Thorium-Plutonium-Actinide Minor (Th-Pu-MA). The utilization of Plutonium in fuel aims to reduce nuclear waste, which is one of the problems in developing nuclear reactors today. Therefore, in this study, an analysis will be carried out on utilizing several types of Plutonium in the MSR 100 MWe. The first analysis of the study was the neutronic analysis of MSR 100 MWe with Th-Pu fuel using several nuclear data libraries (JENDL 3.3, JENDL 4.0, and JEFF 3.1), with the results indicated that these differences affected the neutronic parameters. For example, in the criticality aspect, the reactor can be operated for 2000 days without refueling with a minimum PuF4 concentration of 0.995%, 0.90%, and 0.87% for JEFF 3.1, JENDL 4.0, and JENDL 3.3 calculations, respectively. Moreover, the second analysis regarding the difference in reactor size and variations of Plutonium grade showed that a larger reactor required a lower PuF4 fraction than a small reactor. In addition, the analysis of Plutonium grade variation, such as Weapon Grade, Reactor Grade, and MOX Grade, showed that with a Weapon Grade fuel, the reactor requires the lowest concentration of PuF4, followed by Reactor Grade and MOX Grade, for a reactor operating period of 2000 days. Therefore, from this analysis, a smaller reactor size can reduce more Plutonium. Then from the comparison of neutronic performance with the Th-U fuel scheme and the Th-Pu-MA scheme showed that the reactor could operate for five years with a minimum concentration of UF4 (Th-U scheme) and PuF4 (Th-Pu-MA scheme) are 7.30% and 3.87%, respectively. In addition, the reactivity swing in the Th-U scheme is higher than the Th-Pu-MA, which will also affect the reactor safety factor. Furthermore, the comparative analysis of the molten salt types, such as FLiBe, FLiNaK, and FNaBe, shows that FLiBe molten salt has advantages in terms of neutron economy. The neutronic calculations carried out using the SRAC program (PIJ and CITATION) were also verified using the MCNP program. The results show that the average percentage difference between the effective multiplication factor (keff) is around 2.23%. The obtained result with the SRAC program is reliable because it also corresponds with a shorter calculation time and less high computer specifications.