OpenMC First Tutorial

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

Table of Contents

• 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

Introduction

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.

1. Modeling a Pin-Cell

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

2. Starting Source and Settings

Explore neutron source definition and simulation parameter setup in OpenMC.

3. Running OpenMC

Execute OpenMC simulations and analyze neutron behavior.

4. Geometry Plotting

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

5. The Model Class

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

6. Basic Tallies in OpenMC

Capture neutron physics data with basic tallies during simulations.

7. Universes

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

8. Lattices

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

9. Final CEFR Assembly Modeling Steps

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

10. Advanced Tallies in OpenMC

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

Acknowledgments

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

For any queries, contact: eealanoca@gmail.com.

Happy coding!

Elmer E. Alanoca C.