Removed unnecessary code for old OpenDeplete from build-depleted.py scripts
This commit is contained in:
Will Boyd
parent
aa2721aaf6
commit
5eb830f355
5 changed files with 38 additions and 162 deletions
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@ -9,47 +9,22 @@ import opendeplete
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from geometry import beavrs, openmc_geometry
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Extract cell materials, temperatures and sab
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for cell in openmc_geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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# Extract initial fuel nuclide densities in units of at/cc
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for cell in fuel_cells:
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densities = cell.fill.get_nuclide_atom_densities()
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materials.initial_density[cell.fill.name] = OrderedDict()
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# Convert atom densities from at/b-cm to at/cc
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for nuclide in densities:
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materials.initial_density[cell.fill.name][nuclide.name] = \
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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max_material_id = 0
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for material in openmc_geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Assign distribmats for each material
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for cell in fuel_cells:
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@ -9,47 +9,22 @@ import opendeplete
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from geometry import beavrs, openmc_geometry
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Extract cell materials, temperatures and sab
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for cell in openmc_geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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# Extract initial fuel nuclide densities in units of at/cc
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for cell in fuel_cells:
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densities = cell.fill.get_nuclide_atom_densities()
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materials.initial_density[cell.fill.name] = OrderedDict()
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# Convert atom densities from at/b-cm to at/cc
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for nuclide in densities:
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materials.initial_density[cell.fill.name][nuclide.name] = \
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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max_material_id = 0
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for material in openmc_geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Assign distribmats for each material
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for cell in fuel_cells:
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@ -9,47 +9,22 @@ import opendeplete
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from geometry import beavrs, openmc_geometry
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Extract cell materials, temperatures and sab
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for cell in openmc_geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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# Extract initial fuel nuclide densities in units of at/cc
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for cell in fuel_cells:
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densities = cell.fill.get_nuclide_atom_densities()
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materials.initial_density[cell.fill.name] = OrderedDict()
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# Convert atom densities from at/b-cm to at/cc
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for nuclide in densities:
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materials.initial_density[cell.fill.name][nuclide.name] = \
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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max_material_id = 0
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for material in openmc_geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Assign distribmats for each material
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for cell in fuel_cells:
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@ -9,47 +9,22 @@ import opendeplete
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from geometry import beavrs, openmc_geometry
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Extract cell materials, temperatures and sab
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for cell in openmc_geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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# Extract initial fuel nuclide densities in units of at/cc
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for cell in fuel_cells:
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densities = cell.fill.get_nuclide_atom_densities()
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materials.initial_density[cell.fill.name] = OrderedDict()
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# Convert atom densities from at/b-cm to at/cc
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for nuclide in densities:
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materials.initial_density[cell.fill.name][nuclide.name] = \
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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max_material_id = 0
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for material in openmc_geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 5.
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Assign distribmats for each material
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for cell in fuel_cells:
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@ -10,17 +10,20 @@ from smr.materials import materials
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from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
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from smr.core import geometry
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 200.
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# Count the number of instances for each cell and material
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geometry.determine_paths()
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# Determine the maximum material ID
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max_material_id = 0
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for material in geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# Extract all cells filled by a fuel material
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fuel_cells = geometry.get_cells_by_name(
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name='(1.6%) (0)', case_sensitive=True)
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@ -29,33 +32,6 @@ fuel_cells.extend(geometry.get_cells_by_name(
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fuel_cells.extend(geometry.get_cells_by_name(
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name='(3.1%) (0)', case_sensitive=True))
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# Extract cell materials, temperatures and sab
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for cell in geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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# Extract initial fuel nuclide densities in units of at/cc
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for cell in fuel_cells:
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densities = cell.fill.get_nuclide_atom_densities()
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materials.initial_density[cell.fill.name] = OrderedDict()
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# Convert atom densities from at/b-cm to at/cc
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for nuclide in densities:
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materials.initial_density[cell.fill.name][nuclide.name] = \
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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max_material_id = 0
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for material in geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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# Fuel rod geometric parameters
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radius = 0.39218
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height = 200.
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# Assign distribmats for each material
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for cell in fuel_cells:
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new_materials = []
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