Development and verification of depletion capabilities in the iMC Monte Carlo code

This paper presents the development, optimization, and verification of a depletion module integrated into the iMC Monte Carlo code. Several techniques are implemented to improve the performance and accuracy of the iMC depletion module. In addition, the nuclide control for the depletion of the molten salt reactors is developed. The performance of the depletion module is rigorously assessed through comprehensive code-to-code comparisons with the pre-validated Monte Carlo code Serpent. The evaluation encompasses three distinct depletion scenarios: a single PWR fuel pin, a single SFR fuel pin, and VERA benchmarks. Furthermore, the analysis extends to a simplified molten salt reactor experiment (MSRE) model, incorporating nuclide removal techniques. Comparisons focus on burnup-dependent infinite multiplication factors (k_inf) and nuclide densities of actinides and fission products. Results demonstrate both the high accuracy and enhanced efficiency of the iMC Monte Carlo code’s depletion module, marking a significant advancement in advanced reactor analysis capabilities.