CAD-Based Geometry Representations for Monte Carlo Fusion Neutronics Methods and CSG vs. DAGMC Performance Tradeoffs in OpenMC
Fusion reactors utilizing deuterium and tritium fuel produce high-energy 14.1 MeV neutrons, necessitating a thorough understanding of their behavior for effective reactor design. Neutron transport codes play a critical role in determining key parameters such as tritium breeding ratio, neutron wall l...
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| Format: | Thesis |
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Massachusetts Institute of Technology
2024
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| Online Access: | https://hdl.handle.net/1721.1/157233 |