Hello,
I want to have homogenized cross sections in a specific region of my mesh (which is some cells that I put into universes)
I have two groups of cells, one which is a group of fuel cells, having fissionable materials inside and the other one which is the one of the cell.
I print different cross sections for the fuel cells. The different cross sections are printing values, except Fission and Chi.
Do you have any ideas of the ideas? Thank you in advance! When I do the fission cross section for the whole universe, the fissions rate aren’t null.
# -------------------------- Geometry --------------------------------
root_cell = []
root_cell += [cell for cell in Group1_cell]
root_cell += [cell for cell in Group2_cell]
fuels_cell = []
fuels_cell += [cell for cell in Group2_cell]
# Create universe cells
all_cell_univ = openmc.Universe()
all_cell_univ.add_cells(root_cell)
fuels_cell_univ = openmc.Universe()
fuels_cell_univ.add_cells(fuels_cell)
#2D cell
cell2D = openmc.Cell(fill=all_cell_univ, region = -s_0001&+s_0002)
fuels2D = openmc.Cell(fill=fuels_cell_univ, region = -s_0001&+s_0002)
# Create different universe
fuels_universe = openmc.Universe(universe_id=2, name='fuels universe')
fuels_universe.add_cell(fuels2D)
# Create root Universe
root_universe = openmc.Universe(universe_id=0, name='root universe')
root_universe.add_cell(cell2D)
# ---------------------- Creation of a model -------------------------
# creating model and setting
model = openmc.Model()
model.materials = materials # set materials
model.geometry = openmc.Geometry(root_universe) # set geometry
# ------------------ OpenMC simulation parameters ----------------------
# set to be same as SERPENT model
batches = 500
inactive = 50
particles = 50000
power = 110e6
# Instantiate a Settings object
settings = openmc.Settings()
settings.batches = batches
settings.inactive = inactive
settings.particles = particles
settings.output = {'tallies': False}
#settings.verbosity = 4
# ---- Create an initial uniform spatial source distribution over fissionable zones ------
bounds = [-100.0, -100.0, 5, 100, 100, 6]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings.source = openmc.Source(space=uniform_dist)
model.settings = settings
# ------------ Create the Energy Groups for the library -------------------
# Instantiate a 18-group EnergyGroups object
group_edges = np.array([0., 1E-02, 1E+01, 20.0e6])
groups = openmc.mgxs.EnergyGroups(group_edges = group_edges)
# ----------- Initialize a 18-group MGXS Library for OpenMC -------------
# Create a "tallies.xml" file for the MGXS Library
tallies = openmc.Tallies()
# ------------------------------- Fuels Cross Sections ------------------------------
total_fuels = openmc.mgxs.TotalXS(domain= fuels_universe, energy_groups = groups)
transport_fuels = openmc.mgxs.TransportXS(domain= fuels_universe, energy_groups = groups)
nutransport_fuels = openmc.mgxs.TransportXS(domain= fuels_universe, energy_groups = groups, nu = True)
absorption_fuels = openmc.mgxs.AbsorptionXS(domain= fuels_universe, energy_groups = groups)
capture_fuels = openmc.mgxs.CaptureXS(domain= fuels_universe, energy_groups = groups)
nufission_fuels = openmc.mgxs.FissionXS(domain= fuels_universe, energy_groups = groups, nu = True)
fission_fuels = openmc.mgxs.FissionXS(domain= fuels_universe, energy_groups = groups)
kappafission_fuels = openmc.mgxs.KappaFissionXS(domain= fuels_universe, energy_groups = groups)
scatter_fuels = openmc.mgxs.ScatterXS(domain= fuels_universe, energy_groups = groups)
scattermatrix_fuels = openmc.mgxs.ScatterMatrixXS(domain= fuels_universe, energy_groups = groups)
nuscattermatrix_fuels = openmc.mgxs.ScatterMatrixXS(domain= fuels_universe, energy_groups = groups)
chi_fuels = openmc.mgxs.Chi(domain= fuels_universe, energy_groups = groups)
chiprompt_fuels = openmc.mgxs.Chi(domain= fuels_universe, energy_groups = groups, prompt = True)
inversevelocity_fuels = openmc.mgxs.InverseVelocity(domain= fuels_universe, energy_groups = groups)
fissionpromptnu_fuels = openmc.mgxs.FissionXS(domain= fuels_universe, energy_groups = groups, prompt = True)
chidelayed_fuels = openmc.mgxs.ChiDelayed(domain= fuels_universe, energy_groups = groups)
beta_fuels = openmc.mgxs.Beta(domain= fuels_universe, energy_groups = groups)
reducedabsorption_fuels = openmc.mgxs.ReducedAbsorptionXS(domain= fuels_universe, energy_groups = groups)
# Creating the fuels tally
tallies += total_fuels.tallies.values()
tallies += transport_fuels.tallies.values()
tallies += nutransport_fuels.tallies.values()
tallies += absorption_fuels.tallies.values()
tallies += capture_fuels.tallies.values()
tallies += nufission_fuels.tallies.values()
tallies += fission_fuels.tallies.values()
tallies += kappafission_fuels.tallies.values()
tallies += scatter_fuels.tallies.values()
tallies += scattermatrix_fuels.tallies.values()
tallies += nuscattermatrix_fuels.tallies.values()
tallies += chi_fuels.tallies.values()
tallies += chiprompt_fuels.tallies.values()
tallies += inversevelocity_fuels.tallies.values()
tallies += fissionpromptnu_fuels.tallies.values()
tallies += chidelayed_fuels.tallies.values()
tallies += beta_fuels.tallies.values()
tallies += reducedabsorption_fuels.tallies.values()
# Creating a model
model.tallies = tallies
tallies.export_to_xml()
geometry = openmc.Geometry(root = root_universe)
geometry.export_to_xml()
model.export_to_xml()
# Run OpenMC
statepoint_filename = model.run(threads=48)
# save to h5 file
sp = openmc.StatePoint(statepoint_filename)
# -------------------------- Load fuels Tally -----------------------------
total_fuels.load_from_statepoint(sp)
transport_fuels.load_from_statepoint(sp)
nutransport_fuels.load_from_statepoint(sp)
absorption_fuels.load_from_statepoint(sp)
capture_fuels.load_from_statepoint(sp)
nufission_fuels.load_from_statepoint(sp)
fission_fuels.load_from_statepoint(sp)
kappafission_fuels.load_from_statepoint(sp)
scatter_fuels.load_from_statepoint(sp)
scattermatrix_fuels.load_from_statepoint(sp)
nuscattermatrix_fuels.load_from_statepoint(sp)
chi_fuels.load_from_statepoint(sp)
chiprompt_fuels.load_from_statepoint(sp)
inversevelocity_fuels.load_from_statepoint(sp)
fissionpromptnu_fuels.load_from_statepoint(sp)
chidelayed_fuels.load_from_statepoint(sp)
beta_fuels.load_from_statepoint(sp)
reducedabsorption_fuels.load_from_statepoint(sp)
total_fuels.print_xs()
transport_fuels.print_xs()
nutransport_fuels.print_xs()
absorption_fuels.print_xs()
capture_fuels.print_xs()
nufission_fuels.print_xs()
fission_fuels.print_xs()
kappafission_fuels.print_xs()
scatter_fuels.print_xs()
scattermatrix_fuels.print_xs()
nuscattermatrix_fuels.print_xs()
chi_fuels.print_xs()
chiprompt_fuels.print_xs()
inversevelocity_fuels.print_xs()
fissionpromptnu_fuels.print_xs()
reducedabsorption_fuels.print_xs()
# ------------------------- Creation of a library ----------------------------
# ------------------------- fuels Library ----------------------------
total_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
transport_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
nutransport_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
absorption_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
capture_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
nufission_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
fission_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
kappafission_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
scatter_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
scattermatrix_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
nuscattermatrix_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
chi_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
chiprompt_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
inversevelocity_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
fissionpromptnu_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
chidelayed_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
beta_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
reducedabsorption_fuels.build_hdf5_store(filename='mgxs.h5', directory='./', append=True)
# --------------------------------------------------------------------