OpenMOC is a relatively new reactor physics code that can solve the neutron transport equation using the method of characteristics (MOC), based on the multi-group and a steady-state form of the neutron transport equation. In this study, calculations and simulations were carried out for a TRIGA Mark II Research Reactor using ENDF/B-VII.1 as the nuclear data library to study and test the accuracy of this code and its method. Calculation of the important neutronic parameters in a nuclear reactor design, namely the effective multiplication factor, was carried out for the configuration of core-133 and core-134 in the TRIGA Mark II Research Reactor. The simulation results were compared with experimental results. The results obtained in this study showed a good agreement between the simulation results, using the method of characteristics through the OpenMOC code, and the experimental results with a difference of 0.07% for core-133 and 0.18% for core-134.

Topics
Data visualization, Nuclear reactors, Neutron transport equation, Fission reactors, Reactor physics, Libraries, Public policy and governance
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