Hi all:
I am having a difficulty measuring ‘current’ through a spherical geometry while using 2 different materials at 2 different regions. I have declared the geometry and defined the materials as following: (I am adding the code as text here)
import openmc
import numpy as np
import pandas as pd
import openmc.geometry
import openmc.stats
openmc.config[‘cross_sections’]=‘/Users/sayedahmed/Downloads/Cross_section/endfb-viii.0-hdf5/cross_sections.xml’
z=10 #first material distance so that removal CS times z is less than 5 and approximately 2
x=120 #detector distance
#Spheres
outer_sphere= openmc.Sphere(r=200, boundary_type=‘vacuum’)
detector_sphere= openmc.Sphere(r=z+x)
first_sphere= openmc.Sphere(r=z)
#regions
inside_outer= -outer_sphere & +detector_sphere #everything between outer_sphere and detector sphere
detector_region= -detector_sphere & +first_sphere #everything between detector sphere and first sphere
additional_material_region= -first_sphere #everything inside first sphere
#Cells
water_cell= openmc.Cell(name=‘Water’, fill=water, region=inside_outer)
detector_cell= openmc.Cell(name=‘Detector’, fill=water, region=detector_region)
first_cell= openmc.Cell(name=‘Shield’, fill=additional_material, region=additional_material_region)
#universe
universe= openmc.Universe(cells=[water_cell,detector_cell,first_cell])
geometry= openmc.Geometry(universe)
geometry.export_to_xml()
a=1.18e6
b=1.03419e-6
energy_dist= openmc.stats.Watt(a,b)
#Neutron source
source = openmc.IndependentSource()
source.space = openmc.stats.Point((0.0,0.0,0.0))
source.angle = openmc.stats.Isotropic()
source.energy = energy_dist
#Settings
settings= openmc.Settings()
settings.source= source
settings.batches=100
settings.inactive=10
settings.particles= 100000
settings.run_mode= ‘fixed source’
settings.export_to_xml()
#Tallies
tallies=openmc.Tallies()
surface_filter= openmc.SurfaceFilter(detector_sphere)
energy_filter= openmc.EnergyFilter([0.0, 0.625])
tally= openmc.Tally(name=‘Thermal fluence’)
tally.filters.append(surface_filter)
tally.filters.append(energy_filter)
tally.scores= [‘current’]
tallies.append(tally)
tallies.export_to_xml()
using this material, neutron source, geometry and tallies the output current I am getting is 0. However, if I change the material from Pt to Pb and double the radius of the first sphere then I get a value of output current. What must I do to be able to measure current for any material??
Any reply from anyone is highly appreciated!!
Regards
Sayed