Multichannel analysis of neutronic and thermal-hydraulic coupled calculation for small and long-life BWR

This research aimed to design a small and long-life boiling water reactor (BWR) with thorium-based fuel by analysing coupled neutronic and thermal-hydraulic core calculations. In previous studies, the use of fuel composition Th-232 and U-233 as well as the use of protactinium as burnable poisons with hexagonal tight lattice fuel cell geometry resulted in the design of a BWR that could be operated for 30 years without the necessity to refuel during that period. This study emphasized on the design of small and long-life BWR with thorium based fuel by neutronic and thermal-hydraulic coupling core calculation analysis. Thermal-hydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. Neutronic calculation was carried out by SRAC code and thermal-hydraulic calculation built by FORTRAN code language. This study using a BWR with a cylinder-type reactor core in multichannel heated analysis. The multichannel analysis of neutronic and thermal-hydraulic calculations were carried out by dividing the reactor core into 28 channels in radial directions. Each channel represented the condition that occurs in each point of the reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature were taken into account to obtain distribution of coolant density, flow rate and pressure drop. The calculations were repeated until the void fraction of coolant reactor reached the convergence condition. Finally the design of the reactor core was obtained with 30 years operation time without refuelling, power of 300 MWt and excess reactivity of 0.9%dk/k. The distribution of power density in radial direction was evenly spread using heterogeneous core reactor concept which has a power peaking factor of 1.2. Output from thermal-hydraulic analysis after convergent void fraction on the reactor core mesh was obtained as follows: the maximum void fraction distribution reached 79.7% in the centreline channel of the core reactor and about 74% at the outer channel with an average value of 41.83%; the inlet coolant temperature was 269°C and reached the temperature of 285°C. Furthermore, the coolant total pressure drop in the reactor core was 243.67 kPa.