Preliminary study of 100 MWe Molten Salt Reactor using Monte Carlo Method
Cici Wulandari1 , Abdul Waris2*, Sidik Permana2, Syeilendra Pramuditya2
1Department of Physics, Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
2Department of Physics and Department of Nuclear Science & Engineering,
Faculty of Mathematics and Natural Sciences,
Institut Teknologi Bandung,INDONESIA
*awaris[at]fi.itb.ac.id
Abstract
A preliminary conceptual design is conducted for 100 MW(electric) of Molten Salt Reactor (MSR) fueled with 233U. Neutronic parameters analysis of 100 MWe MSR are carried out by using the Monte Carlo method with MCNP6 program. The reactor can be operated for about 5 years without refueling and any exchange of graphite moderator. The unique liquid salt fuel form increases the reactor proliferation resistance because all radioactive isotopes in the core are hard to be separated. This one is a characteristic of Generation IV nuclear power system. The chemical compositions of fuel salt are LiF-BeF2-ThF4-233UF4. The Thorium utilization in the fuel provided sustainable energy in the reactor due to the breeding capability. The 233U concentration was varied to get the reactor criticality changes. The result showed that 233U loaded in fuel concentration is about 0.9 %mol for burnup 6.33GWd/MTU. This calculation is expected to be a preliminary analysis of the MSR development for future energy.
Keywords: MCNP6, Monte Carlo, MSR, Thorium, Uranium
Topic: Nuclear and Radiation Computation