Selecting Burnup Algorithms in OpenMC Using the Calculated Benchmark of LEU Assembly and MOX Fuel
OpenMC is a state-of-the-art Monte Carlo neutron transport simulation code that uses the Python programming language as an API. OpenMC supports eight burnout simulation algorithms. This study presents the results of choosing an integration method for modeling the burnup of fuel assemblies with burna...
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| Main Authors: | , , , , , |
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| Format: | Article |
| Language: | English |
| Published: |
Национальный исследовательский ядерный университет МИФИ
2023-04-01
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| Series: | Глобальная ядерная безопасность |
| Subjects: | |
| Online Access: | https://glonucsec.elpub.ru/jour/article/view/178 |
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