This repository contains a beginner-friendly tutorial for OpenMC, a Monte Carlo particle transport simulation code widely used in nuclear engineering.

• OpenMC First Tutorial
  • Table of Contents
  • Introduction
  • 1. Modeling a Pin-Cell
  • 2. Starting Source and Settings
  • 3. Running OpenMC
  • 4. Geometry Plotting
  • 5. The Model Class
  • 6. Basic Tallies in OpenMC
  • 7. Universes
  • 8. Lattices
  • 9. Final CEFR Assembly Modeling Steps
  • 10. Advanced Tallies in OpenMC
  • Acknowledgments

This tutorial provides a step-by-step guide to OpenMC, enhancing your understanding of Monte Carlo simulations in nuclear engineering. Special thanks to ICTP-IAEA for their contributions.

Learn how to model a pin-cell, a fundamental step in realistic nuclear reactor simulations.

Explore neutron source definition and simulation parameter setup in OpenMC.

Execute OpenMC simulations and analyze neutron behavior.

Visualize simulation geometry using OpenMC's plotting tools for model verification.

Understand the significance of OpenMC's Model class in reactor model definition.

Capture neutron physics data with basic tallies during simulations.

Discover the concept of universes in OpenMC, crucial for spatial material arrangement.

Efficiently model repetitive structures in reactor cores using OpenMC's lattices.

Conclude the tutorial by modeling a complete reactor assembly with emphasis on the reactor core assembly.

Explore advanced tallies, focusing on obtaining detailed pincell power distributions for comprehensive analysis.

Special thanks to ICTP-IAEA for their invaluable contributions to OpenMC's development.

For any queries, contact: [email protected].

Happy coding!

Elmer E. Alanoca C.