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Fix volume assignment to clad/gap, use proper assemblies

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This commit is contained in:
Paul Romano 2021-08-09 12:40:55 -05:00
parent 1713a4df1b
commit c9554e003a
1 changed files with 17 additions and 8 deletions
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25
smr/build-core-short.py
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@ -66,7 +66,8 @@ fuel_mats = {}
for cell in geometry.get_all_cells().values():
if cell.fill in materials:
# Determine volume of each fuel material
if 'UO2 Fuel' in cell.fill.name:
name = cell.fill.name
if 'UO2 Fuel' in name:
upper_right = cell.region.bounding_box[1][0]
if isclose(upper_right, ring_radii[0]):
ri, ro = 0.0, ring_radii[0]
@ -74,16 +75,17 @@ for cell in geometry.get_all_cells().values():
ri, ro = ring_radii[0], ring_radii[1]
else:
ri, ro = ring_radii[1], pellet_OR
if ri not in fuel_mats:
if (name, ri) not in fuel_mats:
cell.fill = cell.fill.clone()
cell.fill.volume = pi * (ro*ro - ri*ri) * h
fuel_mats[ri] = cell.fill
fuel_mats[name, ri] = cell.fill
else:
cell.fill = fuel_mats[ri]
elif cell.fill.name == 'Helium':
cell.fill = fuel_mats[name, ri]
elif name == 'Helium':
cell.fill.volume = pi * (clad_IR**2 - pellet_OR**2) * h
elif cell.fill.name == 'M5':
cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * h
elif name == 'M5':
# Clad is not subdivided
cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * length
else:
cell.fill.volume = 1.0
@ -93,7 +95,6 @@ all_materials = geometry.get_all_materials()
materials = openmc.Materials(all_materials.values())
materials.export_to_xml(str(directory / 'materials.xml'))
#### Create OpenMC "geometry.xml" file
geometry.export_to_xml(str(directory / 'geometry.xml'))
@ -125,3 +126,11 @@ if args.multipole:
settings.export_to_xml(str(directory / 'settings.xml'))
# Check assembly power distribution
core_lattice = geometry.get_cells_by_fill_name('Main core')[0].fill
mesh = openmc.RegularMesh.from_rect_lattice(core_lattice)
assembly_power = openmc.Tally()
assembly_power.filters = [openmc.MeshFilter(mesh)]
assembly_power.scores = ['nu-fission']
tallies = openmc.Tallies([assembly_power])
tallies.export_to_xml(directory / 'tallies.xml')