This paper describes a newly developed Monte Carlo code used for reactor analysis called RMC1.0, which is based on ACE format library. RMC1.0 is able to estimate criticality eigenvalue, and tally flux/spectrum with collision estimation method or tracking length method. Series of benchmarks and other examples are calculated for validation, which prove that RMC1.0 gives a good performance in both accuracy and efficiency compared with mcnp5. Despite its limitation in geometry processing, RMC1.0 has made a profitable attempt in self-development of Monte Carlo code for reactor analysis.
- Nuclear Engineering Division
RMC1.0: Development of Monte Carlo Code for Reactor Analysis
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She, D, Xu, Q, Wang, K, & Yu, G. "RMC1.0: Development of Monte Carlo Code for Reactor Analysis." Proceedings of the 18th International Conference on Nuclear Engineering. 18th International Conference on Nuclear Engineering: Volume 2. Xi’an, China. May 17–21, 2010. pp. 127-131. ASME. https://doi.org/10.1115/ICONE18-29482
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