Hello, everyone!
I wanted to make sure if i m understanding source defining correctly.
I’m modeling a pin-based reactor core that contains both standard fuel pins and a few designated neutron source pins (filled with UBe13). For my fixed-source simulation, my goal is to ensure that neutrons are born only inside these specific source pins. Here is my code:
bounds = 100
H_core = 60
source_box = openmc.stats.Box(lower_left=(-bounds, -bounds, -H_core),upper_right=(+bounds, +bounds, +H_core), only_fissionable=False)
source_material_filter = openmc.MaterialFilter(UBe13)
source = openmc.IndependentSource()
source.space = source_box
source.energy = source_energy
source.angle = openmc.stats.Isotropic()
source.strength = strength
source.filters = [source_material_filter]
Thank you!