Evaluating the Impact of Neutron Absorber Materials on Spent Fuel Pool Rack Criticality Using OpenMC Monte Carlo Code

Abstract The impact of different neutron absorber materials on the criticality of the spent fuel pool rack geometry has been investigated in this study. OpenMC Monte Carlo code has been used for modelling the geometry of the spent fuel pool rack and the k-eigenvalue/criticality calculations. The criticality calculations consist of three stages. In the first stage, the k-eigenvalue criticality calculations were made using two different neutron absorber plates, which are an aluminium-boron carbide composite and a Gd-containing aluminium-boron carbide composite. In the following stage, k-eigenvalue calculations were performed by removing neutron absorber plates from the rack geometry and adding Gd (1 wt.%) directly to the rack wall composition. In the final stage, criticality calculations were made by using different neutron absorber elements like Gd, B, and Sm and their different concentrations within the rack wall composite. It can be understood from the analyses that Gd is a significant neutron absorber material, particularly when it is incorporated directly within the spent fuel pool rack wall composition. With this approach, the need for neutron absorber plates has been eliminated by adding neutron absorber elements directly within the rack wall composition. In this way, systems will become simpler, and potentially, the cost will be reduced.