Validation of Two-Step Core Kinetics Calculation using Direct Whole Core Transport Solutions, invited

摘要

The current industry practice to perform spatial kinetics calculations for power reactors is to employ the two-step calculation procedure that involves pre-generation of assembly-homogenized few-group cross sections which are then used in the time-dependent diffusion theory-based core calculation. In such transient core calculations, all the fuel pins and flow channels in an assembly are lumped and the thermal feedback is applied on an assembly average basis. Although the intra-pellet fuel temperature distribution is calculated for the assembly averaged pin, the Doppler effect is approximated using an effective fuel temperature which is usually obtained as a weighted average of the centerline and pellet surface temperature. In the spatial kinetics codes in which the two-group formulation is employed mostly, the softer delayed neutron spectrum is treated using an effective delayed neutron fraction.