OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It is capable of performing fixed source, k-eigenvalue and subcritical multiplication calculations on models built using either a constructive solid geometry or CAD representation. OpenMC supports both continuous-energy and multigroup transport. The continuous-energy particle interaction data is based on a native HDF5 format that can be generated from ACE files produced by NJOY. Parallelism is enabled via a hybrid MPI and OpenMP programming model.
More information about the OpenMC Monte Carlo Code.
This four-day interactive course will cover beginner and intermediate fuctions of OpenMC. Topics will include basic model definition, universes and repeated geometry, tallies and post-processing, source definitions, visualisation, depletion, multigroup cross section generation, working with nuclear data, the Python / C / C++ API and multiphysics, variance reduction, and CAD-based geometries. During the course, participants will have access to a pre-installed version of OpenMC on cloud-based servers.
The OpenMC course will be taught in English during the week of 24-27 October 2022 from 10:00-14:00 CT (17:00-21:00 CET).
Maximum number of participants: 35
Course fee: 450 EUR
For any questions about this course, please reach out to Ms. Hedvig Nahon (email@example.com).