BackgroundFusion reactor engineering models are extremely complex, which makes the modeling of the neutronics analysis rather tedious and time consuming. The open-source Monte Carlo code OpenMC can directly perform particle transport simulation on CAD models by integrating DAGMC (Direct Accelerated Geometry Monte Carlo), which can significantly improve the modeling and analysis efficiency of complex engineering models.PurposeThis study aims to carry out the application research of OpenMC in fusion neutronics analysis for China Fusion Engineering Test Reactor (CFETR).MethodsBased on the CFETR 1D cylindrical shell model, the consistency of tallies of OpenMC and MCNP was verified. According to the characteristics of plasma spatial distribution, source class and source function were customized on base of the source extension interface to accurately describe the complex fusion neutron source. The 3D model of complex CFETR engineering was successfully established by using CAD geometry function of DAG-OpenMC, and the neutron wall load distribution, tritium breeding ratio (TBR) and heat were obtained.ResultsThe calculation results of DAG-OpenMC and MCNP have excellent consistency. The neutron wall load peak of CFETR appears at the mid plane, and the tritium breeding ratio of helium cooled ceramic breeder blanket is 1.141. When the fusion neutron power is 160 MW, the total nuclear heat generated is 211.73 MW, and the global energy multiplication factor is 1.32, in which the total nuclear heat generated by the blanket is 185.64 MW, accounting for 87.68% of the total nuclear heat.ConclusionsThe CAD-based geometry capabilities greatly improve the modeling efficiency of building complex fusion reactor engineering models. Verification of the key issues of DAG-OpenMC in fusion neutronics applications by this study, indicates its feasibility in dealing with fusion neutronics problems under complex engineering structural conditions.