Burnup computation for HTR-10 using MCNPX as the function of radius and fuel enrichment

Abstract

Abstract The HTR-10 is a gas-cooled high-temperature reactor with spherical fuel and moderator called pebble bed and operated on 10 MW thermal power. Until now, a lot of research and development on the HTR-10 in terms of neutronic computational modeling has been carried out, one of which is burnup analysis. Fuel depletion or burnup analysis is an analysis related to fuel control and reactor output power. This analysis is needed because changes in fuel composition will affect the neutronic parameter values of a reactor. The 10 MW HTR fuel depletion study has been conducted by modeling the reactor using MCNPX (Monte Carlo N-Particle e-Xtended) to analyze the effect of radius and enrichment of reactor fuel to the left value and the burnup rate. The study is conducted by modeling a 10 MW HTR with three sorts of fuel kernel radiuses using MCNPX. First, 250 μm radius, second, 275 μm radius, third, 300 μm radius and all radiuses are variated with 10-15% fuel enrichment. TRISO particles are dispersed using simple cubic (SC) lattice in the fuel zone. The fuel balls are allocated in the reactor core along with moderator balls using a body-centered cubic. To perform all the calculations, MCNPX utilizes continuous energy data library ENDF/B-VI. From the calculation result using MCNPX can be concluded that the kernels with the radius of 275 μm and 300 μm at 14% and 13% fuel enrichment respectively were more able to maintain a critical condition within a year of operation than the kernels a 250 μm radius and the fuel with 250 μm radius has the highest burnup rate that is 52,07 GWd/MTU.