Problem Description
I’m encountering an issue with a simple shielding calculation for a fusion neutron source. I’m sampling the thickness of the shielding layer and filtering in a material_tissue defined as an operator. The shielding is being compared with two different materials: one is simply water, and the other is high-density borated polyethylene.
Sampling works fine, but when I attempt to run a simulation for each material with a larger number of particles to visualize the dose map, I face a problem. The simulation works correctly for high-density borated polyethylene, but for water, it gets stuck in a specific batch depending on the seed defined.
One brute-force solution could be trying different seeds until it works or running simulations with a low number of batches and then combining the results. However, I find these approaches somewhat archaic.
Talking with @Shimwell I’ve tried recompiling OpenMC with the suggestion provided in this link, but it still doesn’t work. Any suggestions on what might be happening would be greatly appreciated.
Environment
- OpenMC version 0.14.1 (I did not encounter this problem with version 0.13.3)
- Cross sections: ENDFB-8.0-NNDC and TENDL-2019 (ENDF with TENDL where ENDF cross sections are not available).
- Cases run both locally (12 CPU, 32 GM RAM) and on a cluster (128 CPU).
- Not likely a particle-per-batch issue, as I’ve run more complex models with many particles successfully.
Material Definitions
Water
water = openmc.Material(name="h2o")
water.add_nuclide("H1", 2.0)
water.add_nuclide("O16", 1.0)
water.set_density('g/cm3', 1.0)
borated_polyethilene = openmc.Material(name="borated_polyethilene")
borated_polyethilene.add_nuclide("C12", 61.20, percent_type='wo')
borated_polyethilene.add_nuclide("H1", 11.60, percent_type='wo')
borated_polyethilene.add_nuclide("B10", 1, percent_type='wo')
borated_polyethilene.add_nuclide("B11", 4, percent_type='wo')
borated_polyethilene.add_nuclide("O16", 22.20, percent_type='wo')
borated_polyethilene.set_density('g/cm3', 0.94)
Model geometry
Pink: Cylindrical material_tissue
Red: Cylindrical Neutron source
Grey: Cylindrical Shielding
White: Vacuum cell with “vacuum” boundary conditions
