Validation of Unstructured Mesh Capability for Modeling Variable Density Objects Using MCNP®

Author

William K. Backman

Date of Award

3-2023

Document Type

Thesis

Degree Name

Master of Science in Nuclear Engineering

Department

Department of Engineering Physics

First Advisor

Darren E. Holland, PhD

Abstract

MCNP® is a radiation transport code with the capability to define geometries using unstructured meshes. While this geometry type was intended to help model complex geometries, it also has the capability of being used to model objects with variable density. This can be done by assigning varying density values to different elements in a mesh. This capability has not been well explored, though it could be beneficial for several applications within the nuclear engineering community.

AFIT Designator

AFIT-ENP-MS-23-M-075

Comments

A 12-month embargo was observed.

Approved for public release. PA case number on file.

Recommended Citation

Backman, William K., "Validation of Unstructured Mesh Capability for Modeling Variable Density Objects Using MCNP®" (2023). Theses and Dissertations. 7342.
https://scholar.afit.edu/etd/7342