def initialize(self, settings):
""" Initializes the geometry.
After geometry, volume, and materials are set, this function completes
the geometry.
Parameters
----------
settings : openmc_wrapper.Settings
Settings to initialize with.
"""
import copy
# Clear out OpenMC
openmc.reset_auto_material_id()
openmc.reset_auto_surface_id()
openmc.reset_auto_cell_id()
openmc.reset_auto_universe_id()
self.number_density = OrderedDict()
self.mat_name = OrderedDict()
self.burn_cell_to_ind = OrderedDict()
self.burn_list = []
self.fet_order = settings.fet_order
cell_ind = 0
# First, for each cell, create an entry in number_density
cells = self.geometry.root_universe.get_all_cells()
#print(cells)
for cid in cells:
cell = cells[cid]
name = cell.name
self.mat_name[cid] = name
# Create entry in self.number_density using initial_density
self.number_density[cid] = \
copy.deepcopy(self.materials.initial_density[name])
# print('testing...',cid)
# Fill with a material (linked to cell id)
cell.fill = self.density_dictionary_to_openmc_mat(cid)
# If should burn, add to burn list:
if self.materials.burn[name]:
self.burn_list.append(cid)
self.burn_cell_to_ind[str(cid)] = cell_ind
cell_ind += 1
# Then, write geometry.xml
self.geometry.export_to_xml()
# Load participating nuclides
self.load_participating(settings.cross_sections)
# Create reaction rate tables
self.reaction_rates = \
reaction_rates.ReactionRates(self.burn_cell_to_ind,
self.burn_nuc_to_ind,
self.chain.react_to_ind,
self.fet_order)
# Finally, calculate total number densities
self.total_number = OrderedDict()
self.calculate_total_number()
def generate_materials_xml(self):
""" Creates materials.xml from self.number_density.
Iterates through each cell in self.number_density and creates the
openmc material object to generate materials.xml.
"""
openmc.reset_auto_material_id()
mat = []
i = 0
for key_mat in self.number_density:
mat.append(openmc.Material(material_id=key_mat))
mat_name = self.mat_name[key_mat]
total = 0.0
for key_nuc in self.number_density[key_mat]:
# Check if in participating nuclides
if key_nuc in self.participating_nuclides:
nuc = openmc.Nuclide(key_nuc,
xs=self.materials.library[mat_name])
if type(self.number_density[key_mat][key_nuc]) is zernike.ZernikePolynomial:
nuc.poly_coeffs = self.number_density[key_mat][key_nuc].openmc_form(radial_only=False)
self.total_number[key_mat][key_nuc] = self.number_density[key_mat][key_nuc] * self.volume[key_mat]
nuc.poly_type = "zernike"
mat[i].add_nuclide(nuc,
self.number_density[key_mat][key_nuc].coeffs[0])
total += self.number_density[key_mat][key_nuc].coeffs[0]
else:
mat[i].add_nuclide(nuc,
self.number_density[key_mat][key_nuc])
total += self.number_density[key_mat][key_nuc]
mat[i].set_density('atom/cm3', total)
if mat_name in self.materials.sab:
mat[i].add_s_alpha_beta(self.materials.sab[mat_name],
self.materials.library_sab[mat_name])
i += 1
materials_file = openmc.Materials()
materials_file.add_materials(mat)
materials_file.export_to_xml()
def reset_auto_ids():
openmc.reset_auto_material_id()
openmc.reset_auto_surface_id()
openmc.reset_auto_cell_id()
openmc.reset_auto_universe_id()
def reset_auto_ids():
openmc.reset_auto_material_id()
openmc.reset_auto_surface_id()
openmc.reset_auto_cell_id()
openmc.reset_auto_universe_id()
def reset_auto_ids():
"""Reset counters for all auto-generated IDs"""
openmc.reset_auto_material_id()
openmc.reset_auto_surface_id()
openmc.reset_auto_cell_id()
openmc.reset_auto_universe_id()