Testing the Activation Analysis for Fusion in OpenMC

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It can perform fission simulations, such as fixed-source,
k
-eigenvalue, and subcritical multiplication calculations on models built using either a constructive solid geometry (CSG) or CAD representation. To explore the use of OpenMC for fusion activation analysis, a detailed model of the Fusion Neutronics Science Facility (FNSF) was first developed for comparisons against an existing SERPENT model. A 90
$^{\circ}$
model of FNSF in standard-triangle language (STL) CAD format was converted to CSG using each code’s built-in functions, and the geometries were validated by ensuring no cells overlapped and no particles were lost during simulations. The neutron fluxes were calculated and compared for multiple components close to the plasma. The results show differences mostly below 1% in fluxes and averaged 8% for activity and decay heat. The work described in this study tests the CAD-based geometry using the DagMC toolkit in OpenMC and compares the activation analysis of OpenMC to SERPENT code.