Hello everone. I would like to how to use OpenMC to calculate the power distribution of assemblies.
Take the CAR-3600MW Core as an example, the core was divided into inner and outer core zones, which are composed of 225 and 228 fuel assemblies, respectively. Each fuel assembly had 469 fuel pins containing five different materials in the axial direction.
I tried using openmc.DistribcellFilter to get the fission reaction rate of each cell and using openmc.MaterialFilter to get the kappa-fission of each materials. However, there was a big problem.
I has fossion results for (225+228)46910 = 2,124,570 distribcell like this.
This cannot be done using human statistics to get the desired results.
How do I complete the data processing or how should I change my tallies?