Learn about OpenMC

[yardasol]

For this project, we will be using OpenMC, an open-source, community developed monte-carlo neutron particle transport code.

Objectives

• Be able to describe the basic program flow of OpenMC
• Understand the concepts of source convergence as it relates to monte carlo k-eigenvalue simulations
• Be able to create OpenMC input files from scratch using the Python API
• Be able to run a simple k-eigenvalue simulation and process the results, and plot the fission reaction rate on the geometry.

Guiding questions

1. What is a Monte Carlo code? How does it differ from "deterministic" codes? What are the advantages and disadvantages of of Monte Carlo codes compared to deterministic codes?
2. What is CSG geometry? How is CSG geometry used to construct physical models in OpenMC?
3. What are the necessary attributes of an openmc.Material object?
4. What is weight percent? How is it different from atom percent? (Hint: Read Chapter 1 in Shultis and Faw, ISBN 1498769292)
5. What is a tally? What is a score? What is a filter? How can we use these three ideas to get quantities of interest in a Monte Carlo code? (Hint: Read section 5 in the MC21 paper)

Resources

• OpenMC Theory Manual (Focus on Sections 1, 2, 3.1, 3.3, 7, and 8)
• OpenMC User's Guide (Focus on Sections 1, 2.4, 3.1, 4, 5, 6.1-6.4, 7.1-7.2, 7.4, 8, 9, and 12)
• Pincell example notebook
• Matplotlib Tutorial