Task Introduction

Please allow 15 minutes for this task.

In this task you will use OpenMC to tally interactions in a geometry using 2D and 3D mesh tallies. Note that unstructured mesh simulations can be found in the CAD part of the workshop.

Learning Outcomes

• Mesh tallies can be used in neutronics simulations to measure a variety of different reactions such as neutron absorption, tritium production and flux.

• How neutrons are dissipated around the reactor.