Modeling and Analysis of Neutron Flux of a VVER 1200 Reactor Core Using Monte Carlo Code OpenMC - K. M. Rakib Al Hasan; Dr. Debashis Datta; Altab Hossain (International Journal of Nuclear Security, 2024)
Nuclear reactors and their associated facilities are complex systems that require accurate modeling and analysis to ensure safety, security, and efficient operation. This study aims to model and analyze the neutron flux of a generic VVER 1200 reactor core using the Monte Carlo code OpenMC. For the purpose of this analysis, a generic model of the VVER 1200 reactor core was developed using an OpenMC simulation environment. Furthermore, using this model, the neutron flux spectrum inside the reactor core was evaluated, from which neutron fluence was calculated accordingly. This study analyzed the effect of neutron flux on the developed model of the VVER 1200 reactor core. The research results reveal a close relationship between neutron flux and reactor power, whereby an escalation in neutron flux leads to a proportional escalation in reactor power output, indicating their interdependence. Additionally, the results highlight a strong correlation between neutron fluence and neutron flux, where an elevation in neutron flux causes a corresponding rise in neutron fluence, which is a significant contributor to the increase in the ductile-to-brittle transition temperature of the RPV wall. In this study, the VVER 1200 reactor core model was simulated and after 60 years of operation; the fluence, which is the total neutron flux received by the reactor vessel, was extracted from the evaluated neutron flux spectrum, and its value was found to be at the design limit of the 1019 scale. So, it is important to maintain the neutron flux at a level that provides sufficient power output while ensuring the RPV’s safety and longevity. The results also show that OpenMC is an effective tool for simulating the neutron flux distribution in the VVER 1200 reactor core. The study findings provide valuable insights into the behavior of the reactor core, which can help improve the design and operation of nuclear reactors for safer and more efficient nuclear energy production.
Keywords: Monte Carlo code OpenMC; VVER 1200; reactor pressure vessel; neutron flux