From 5eb830f355b1cd02f6def3257fe5372f9d21e44a Mon Sep 17 00:00:00 2001 From: Will Boyd Date: Fri, 10 Mar 2017 12:23:58 -0500 Subject: [PATCH] Removed unnecessary code for old OpenDeplete from build-depleted.py scripts --- 2x2-periodic/build-depleted.py | 39 ++++++----------------------- 2x2-reflector/build-depleted.py | 39 ++++++----------------------- assembly/build-depleted.py | 39 ++++++----------------------- pin-cell/build-depleted.py | 39 ++++++----------------------- smr/build-depleted.py | 44 ++++++++------------------------- 5 files changed, 38 insertions(+), 162 deletions(-) diff --git a/2x2-periodic/build-depleted.py b/2x2-periodic/build-depleted.py index 847dca7..3487baa 100644 --- a/2x2-periodic/build-depleted.py +++ b/2x2-periodic/build-depleted.py @@ -9,47 +9,22 @@ import opendeplete from geometry import beavrs, openmc_geometry -# Setup OpenDeplete Materials wrapper -materials = opendeplete.Materials() -materials.temperature = OrderedDict() -materials.sab = OrderedDict() -materials.initial_density = OrderedDict() -materials.burn = OrderedDict() -materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"] +# FIXME: Automatically extract info needed to calculate burnable cell volumes +# Fuel rod geometric parameters +radius = 0.39218 +height = 5. # Count the number of instances for each cell and material openmc_geometry.determine_paths() -# Extract all cells filled by a fuel material -fuel_cells = openmc_geometry.get_cells_by_name( - name='enr radial 0: Fuel', case_sensitive=True) - -# Extract cell materials, temperatures and sab -for cell in openmc_geometry.get_all_material_cells().values(): - materials.burn[cell.name] = 'fuel' in cell.fill.name.lower() - materials.temperature[cell.name] = 300 - if len(cell.fill._sab) > 0: - materials.sab[cell.name] = cell.fill._sab[0] - -# Extract initial fuel nuclide densities in units of at/cc -for cell in fuel_cells: - densities = cell.fill.get_nuclide_atom_densities() - materials.initial_density[cell.fill.name] = OrderedDict() - - # Convert atom densities from at/b-cm to at/cc - for nuclide in densities: - materials.initial_density[cell.fill.name][nuclide.name] = \ - densities[nuclide][1] * 1e24 - # Determine the maximum material ID max_material_id = 0 for material in openmc_geometry.get_all_materials().values(): max_material_id = max(max_material_id, material.id) -# FIXME: Automatically extract info needed to calculate burnable cell volumes -# Fuel rod geometric parameters -radius = 0.39218 -height = 5. +# Extract all cells filled by a fuel material +fuel_cells = openmc_geometry.get_cells_by_name( + name='enr radial 0: Fuel', case_sensitive=True) # Assign distribmats for each material for cell in fuel_cells: diff --git a/2x2-reflector/build-depleted.py b/2x2-reflector/build-depleted.py index b116a84..ce76884 100644 --- a/2x2-reflector/build-depleted.py +++ b/2x2-reflector/build-depleted.py @@ -9,47 +9,22 @@ import opendeplete from geometry import beavrs, openmc_geometry -# Setup OpenDeplete Materials wrapper -materials = opendeplete.Materials() -materials.temperature = OrderedDict() -materials.sab = OrderedDict() -materials.initial_density = OrderedDict() -materials.burn = OrderedDict() -materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"] +# FIXME: Automatically extract info needed to calculate burnable cell volumes +# Fuel rod geometric parameters +radius = 0.39218 +height = 5. # Count the number of instances for each cell and material openmc_geometry.determine_paths() -# Extract all cells filled by a fuel material -fuel_cells = openmc_geometry.get_cells_by_name( - name='enr radial 0: Fuel', case_sensitive=True) - -# Extract cell materials, temperatures and sab -for cell in openmc_geometry.get_all_material_cells().values(): - materials.burn[cell.name] = 'fuel' in cell.fill.name.lower() - materials.temperature[cell.name] = 300 - if len(cell.fill._sab) > 0: - materials.sab[cell.name] = cell.fill._sab[0] - -# Extract initial fuel nuclide densities in units of at/cc -for cell in fuel_cells: - densities = cell.fill.get_nuclide_atom_densities() - materials.initial_density[cell.fill.name] = OrderedDict() - - # Convert atom densities from at/b-cm to at/cc - for nuclide in densities: - materials.initial_density[cell.fill.name][nuclide.name] = \ - densities[nuclide][1] * 1e24 - # Determine the maximum material ID max_material_id = 0 for material in openmc_geometry.get_all_materials().values(): max_material_id = max(max_material_id, material.id) -# FIXME: Automatically extract info needed to calculate burnable cell volumes -# Fuel rod geometric parameters -radius = 0.39218 -height = 5. +# Extract all cells filled by a fuel material +fuel_cells = openmc_geometry.get_cells_by_name( + name='enr radial 0: Fuel', case_sensitive=True) # Assign distribmats for each material for cell in fuel_cells: diff --git a/assembly/build-depleted.py b/assembly/build-depleted.py index 6d36f28..4d42908 100644 --- a/assembly/build-depleted.py +++ b/assembly/build-depleted.py @@ -9,47 +9,22 @@ import opendeplete from geometry import beavrs, openmc_geometry -# Setup OpenDeplete Materials wrapper -materials = opendeplete.Materials() -materials.temperature = OrderedDict() -materials.sab = OrderedDict() -materials.initial_density = OrderedDict() -materials.burn = OrderedDict() -materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"] +# FIXME: Automatically extract info needed to calculate burnable cell volumes +# Fuel rod geometric parameters +radius = 0.39218 +height = 5. # Count the number of instances for each cell and material openmc_geometry.determine_paths() -# Extract all cells filled by a fuel material -fuel_cells = openmc_geometry.get_cells_by_name( - name='enr radial 0: Fuel', case_sensitive=True) - -# Extract cell materials, temperatures and sab -for cell in openmc_geometry.get_all_material_cells().values(): - materials.burn[cell.name] = 'fuel' in cell.fill.name.lower() - materials.temperature[cell.name] = 300 - if len(cell.fill._sab) > 0: - materials.sab[cell.name] = cell.fill._sab[0] - -# Extract initial fuel nuclide densities in units of at/cc -for cell in fuel_cells: - densities = cell.fill.get_nuclide_atom_densities() - materials.initial_density[cell.fill.name] = OrderedDict() - - # Convert atom densities from at/b-cm to at/cc - for nuclide in densities: - materials.initial_density[cell.fill.name][nuclide.name] = \ - densities[nuclide][1] * 1e24 - # Determine the maximum material ID max_material_id = 0 for material in openmc_geometry.get_all_materials().values(): max_material_id = max(max_material_id, material.id) -# FIXME: Automatically extract info needed to calculate burnable cell volumes -# Fuel rod geometric parameters -radius = 0.39218 -height = 5. +# Extract all cells filled by a fuel material +fuel_cells = openmc_geometry.get_cells_by_name( + name='enr radial 0: Fuel', case_sensitive=True) # Assign distribmats for each material for cell in fuel_cells: diff --git a/pin-cell/build-depleted.py b/pin-cell/build-depleted.py index 4ee34b6..dc8b23e 100644 --- a/pin-cell/build-depleted.py +++ b/pin-cell/build-depleted.py @@ -9,47 +9,22 @@ import opendeplete from geometry import beavrs, openmc_geometry -# Setup OpenDeplete Materials wrapper -materials = opendeplete.Materials() -materials.temperature = OrderedDict() -materials.sab = OrderedDict() -materials.initial_density = OrderedDict() -materials.burn = OrderedDict() -materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"] +# FIXME: Automatically extract info needed to calculate burnable cell volumes +# Fuel rod geometric parameters +radius = 0.39218 +height = 5. # Count the number of instances for each cell and material openmc_geometry.determine_paths() -# Extract all cells filled by a fuel material -fuel_cells = openmc_geometry.get_cells_by_name( - name='enr radial 0: Fuel', case_sensitive=True) - -# Extract cell materials, temperatures and sab -for cell in openmc_geometry.get_all_material_cells().values(): - materials.burn[cell.name] = 'fuel' in cell.fill.name.lower() - materials.temperature[cell.name] = 300 - if len(cell.fill._sab) > 0: - materials.sab[cell.name] = cell.fill._sab[0] - -# Extract initial fuel nuclide densities in units of at/cc -for cell in fuel_cells: - densities = cell.fill.get_nuclide_atom_densities() - materials.initial_density[cell.fill.name] = OrderedDict() - - # Convert atom densities from at/b-cm to at/cc - for nuclide in densities: - materials.initial_density[cell.fill.name][nuclide.name] = \ - densities[nuclide][1] * 1e24 - # Determine the maximum material ID max_material_id = 0 for material in openmc_geometry.get_all_materials().values(): max_material_id = max(max_material_id, material.id) -# FIXME: Automatically extract info needed to calculate burnable cell volumes -# Fuel rod geometric parameters -radius = 0.39218 -height = 5. +# Extract all cells filled by a fuel material +fuel_cells = openmc_geometry.get_cells_by_name( + name='enr radial 0: Fuel', case_sensitive=True) # Assign distribmats for each material for cell in fuel_cells: diff --git a/smr/build-depleted.py b/smr/build-depleted.py index 84af9bf..1fa0846 100644 --- a/smr/build-depleted.py +++ b/smr/build-depleted.py @@ -10,17 +10,20 @@ from smr.materials import materials from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core from smr.core import geometry -# Setup OpenDeplete Materials wrapper -materials = opendeplete.Materials() -materials.temperature = OrderedDict() -materials.sab = OrderedDict() -materials.initial_density = OrderedDict() -materials.burn = OrderedDict() -materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"] + +# FIXME: Automatically extract info needed to calculate burnable cell volumes +# Fuel rod geometric parameters +radius = 0.39218 +height = 200. # Count the number of instances for each cell and material geometry.determine_paths() +# Determine the maximum material ID +max_material_id = 0 +for material in geometry.get_all_materials().values(): + max_material_id = max(max_material_id, material.id) + # Extract all cells filled by a fuel material fuel_cells = geometry.get_cells_by_name( name='(1.6%) (0)', case_sensitive=True) @@ -29,33 +32,6 @@ fuel_cells.extend(geometry.get_cells_by_name( fuel_cells.extend(geometry.get_cells_by_name( name='(3.1%) (0)', case_sensitive=True)) -# Extract cell materials, temperatures and sab -for cell in geometry.get_all_material_cells().values(): - materials.burn[cell.name] = 'fuel' in cell.fill.name.lower() - materials.temperature[cell.name] = 300 - if len(cell.fill._sab) > 0: - materials.sab[cell.name] = cell.fill._sab[0] - -# Extract initial fuel nuclide densities in units of at/cc -for cell in fuel_cells: - densities = cell.fill.get_nuclide_atom_densities() - materials.initial_density[cell.fill.name] = OrderedDict() - - # Convert atom densities from at/b-cm to at/cc - for nuclide in densities: - materials.initial_density[cell.fill.name][nuclide.name] = \ - densities[nuclide][1] * 1e24 - -# Determine the maximum material ID -max_material_id = 0 -for material in geometry.get_all_materials().values(): - max_material_id = max(max_material_id, material.id) - -# FIXME: Automatically extract info needed to calculate burnable cell volumes -# Fuel rod geometric parameters -radius = 0.39218 -height = 200. - # Assign distribmats for each material for cell in fuel_cells: new_materials = []