3-Dimensional full core neutronic analysis for uranium nitride fuel CANDLE reactor using Monte Carlo MCNP6 code

The neutronic analysis of full-core 3-dimensional geometrical modeling for a CANDLE reactor was conducted in
this study, utilizing uranium nitride fuel and the MCNP6 code based on continuous energy. The fuel core
composition was sourced from advanced pressurized water reactor (APWR) spent fuel with a 5-year cooling time
and 49 GWd/t of burnup. The burnable poison adjustment was omitted in this design, as theoretically, the
CANDLE reactor can maintain criticality without control rods. To streamline the modeling process, the radial
direction of the fuel core composition was homogeneously designed. Additionally, for tracking the burning wave
and assessing neutronic performance, the axial direction was segmented into 10 regions, each with a 25-cm
height. The total burnup calculation spanned 80 years, with the assumption of 10,000 neutron number his
tories per cycle, aimed at identifying the time required to reach equilibrium. This study also investigated the
impact of thermal power variation on neutronic results within the same core geometry. It was observed that
power distribution and nuclide number densities shifted with a constant shape in the axial direction. As a breeder
reactor, the conversion ratio of the CANDLE reactor for various thermal powers exceeded unity, with higher
thermal power facilitating faster equilibrium state attainment. Burning wave velocity was calculated by
analyzing the shifting of power peak positions for different thermal power core designs, revealing velocities of
approximately ~ 2.5 cm/year, ~1.56 cm/year, and ~ 1 cm/year for thermal powers of 800 MWth, 600 MWth,
and 400 MWth, respectively.