This workshop will provide an introduction to the fundamentals of fusion neutronics through a combination of theory and simulation exercises to be carried out with OpenMC. OpenMC is a general purpose Monte Carlo neutron and photon transport simulation code. It is capable of simulating 3D models based on constructive solid geometry as well as CAD-based geometries using the DAGMC library. OpenMC was originally developed by members of the Computational Reactor Physics Group at the Massachusetts Institute of Technology starting in 2011 with a specific focus on high performance computing and has now evolved into a community developed code with contributions from many institutions. The specific programming and analysis exercises that will be explored in the workshop include: neutron and photon transport, activation simulations, tritium production, nuclear heating, and shutdown dose rates. All levels of experience with fusion neutronics are welcome. Our only recommendation is that attendees have some experience with the Python programming language.