Neutronics analysis is a foundational component of fusion energy research, informing studies of shielding, tritium breeding, and component performance. This tutorial provides a highly compressed introduction to fusion neutronics modeling using OpenMC, an open-source Monte Carlo particle transport code. This short tutorial emphasizes core concepts and workflows, including problem setup, fusion-relevant neutron source definition, and the extraction of representative neutronic quantities. Given the limited time available, a small subset of OpenMC’s capabilities will be explored with the goal of providing attendees a conceptual roadmap for applying OpenMC to fusion problems and a clear sense of where to go next for deeper, more realistic studies.