The Moroccan TRIGA Mark II, a 2-MW(thermal) open-pool light water reactor, has been extensively studied using various computational codes; however, an efficient and accurate transport-depletion coupling model has not yet been developed. This study aims to fill this gap by developing a coupled transport-depletion model based on the open-source Monte Carlo code OpenMC Python API and comparing its performance with the MCNP6.2 code using its burn capability, CINDER90. Two depletion chains were employed in the model: the full depletion chain, which includes all relevant isotopes for higher accuracy, and the reduced depletion chain, which focuses on the most significant isotopes to balance precision and computational efficiency. The key parameters analyzed included the effective multiplication factor (keff), fuel composition changes (both overall and by ring to understand their spatial distribution), neutron spectrum, and flux distribution.

The results demonstrated strong agreement between OpenMC and MCNP6.2, confirming the accuracy and reliability of the new model. The keff values from both depletion chains were consistent with MCNP6.2, and the actinide buildup results aligned well with both codes. While the fission product buildup showed good alignment for most isotopes, discrepancies in 85Kr and 109Ag were observed due to different handling methodologies for fission product yields (FPYs), emphasizing the importance of using either independent or cumulative FPYs. A neutron spectrum analysis indicated that burnup primarily affects the thermal neutron domain, with increased thermal flux as fissile material decreases, and significant changes in the inner rings of the reactor core due to higher burnup levels.