Hi all so I am doing this shielding callculation for a reactor and I have defiend a cylindrical mesh over region of interest where I want to calculate the flux
TALLIES
#energy_bins (ev)
energy_filter_n = openmc.EnergyFilter([0, 5e5, 2e6])
energy_filter_p =openmc.EnergyFilter([5e5,1.5e6,2.5e6,4e6,6e6,8e6,10e6])
tallies = openmc.Tallies()
# Radial and axial mesh edges (e.g. 11 bins each, 10 cm wide)
r_grid = np.linspace(88.45, 99.45, num=11) # 10 radial bins
z_grid = np.linspace(-50, 50, num=11) # 10 axial bins
# Collapse phi into a single bin (you can also add more if needed)
phi_grid = [0.0, 2*np.pi] # Just one full circle
# Construct the cylindrical mesh
mesh = openmc.CylindricalMesh(r_grid=r_grid, z_grid=z_grid, phi_grid=phi_grid)
# Mesh flux tally for neutrons
mesh_filter = openmc.MeshFilter(mesh)
tally_flux_n = openmc.Tally(tally_id=7, name=‘Detector_Neutron_Flux’)
tally_flux_n.filters = [mesh_filter, energy_filter_n]
tally_flux_n.scores = [‘flux’]
tally_flux_n.particles = [‘neutron’]
tallies.append(tally_flux_n)
now I want to normalize this flux into usual units. Ik I have to multiply it by source strength and divide by mesh volume for each mesh bin, but I dont know how to do it . cna someone help in providing a post processing script for this.