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Updated scripts per comments by @paulromano

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This commit is contained in:
Will Boyd 2017-03-18 21:42:59 -04:00
parent 736c9d6d0c
commit 827efd3cf2
10 changed files with 34 additions and 64 deletions
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8
2x2-periodic/build-fresh.py
View file

@ -21,7 +21,7 @@ openmc_geometry.export_to_xml()
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = True if multipole == 'y' else False
multipole = (multipole == 'y')
# Construct uniform initial source distribution over fissionable zones
lower_left = [-21.41728, -21.41728, +192.5]
@ -52,9 +52,9 @@ beavrs.write_openmc_plots()
plot = openmc.Plot()
plot.width = [21.41728*2, 21.41728*2]
plot.origin = [0., 0., 195.]
plot.color = 'mat'
plot.color_by = 'material'
plot.filename = '2x2-periodic'
plot.col_spec = beavrs.plots.colspec_mat
plot.colors = beavrs.plots.colors_mat
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
@ -106,7 +106,7 @@ mesh.type = 'regular'
mesh.dimension = [34, 34, 1]
mesh.lower_left = lower_left
mesh.width = (np.array(upper_right) - np.array(lower_left))
mesh.width[:2] /= 34
mesh.width[:2] /= mesh.dimension
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate energy-integrated fission rate mesh Tally

21
2x2-periodic/geometry.py
View file

@ -44,20 +44,15 @@ def find_assembly(assembly_name, wrap_geometry=True):
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.add_surface(surface=min_z, halfspace=+1)
root_cell.add_surface(surface=max_z, halfspace=-1)
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry()
fuel_assembly.root_universe = root_univ
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly
@ -122,20 +117,14 @@ def build_two_by_two(assembly1_name, assembly2_name):
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.add_surface(surface=min_z, halfspace=+1)
root_cell.add_surface(surface=max_z, halfspace=-1)
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe for this fuel assembly
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create an OpenMC Geometry for this fuel assembly
two_by_two = openmc.Geometry()
two_by_two.root_universe = root_univ
two_by_two = openmc.Geometry(root_univ)
return two_by_two

8
2x2-reflector/build-fresh.py
View file

@ -21,7 +21,7 @@ openmc_geometry.export_to_xml()
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = True if multipole == 'y' else False
multipole = (multipole == 'y')
# Construct uniform initial source distribution over fissionable zones
lower_left = [-32.12592, -32.12592, 192.5]
@ -55,9 +55,9 @@ beavrs.write_openmc_plots()
plot = openmc.Plot()
plot.width = [64.25184, 64.25184]
plot.origin = [0., 0., 195.]
plot.color = 'mat'
plot.color_by = 'material'
plot.filename = '2x2-reflector'
plot.col_spec = beavrs.plots.colspec_mat
plot.colors = beavrs.plots.colors_mat
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
@ -109,7 +109,7 @@ mesh.type = 'regular'
mesh.dimension = [34, 34, 1]
mesh.lower_left = [lower_left[0], lower_left[1] + lat_width[1], lower_left[2]]
mesh.width = np.array(lat_width)
mesh.width[:2] /= 17.
mesh.width[:2] /= mesh.dimension
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate energy-integrated fission rate mesh Tally

18
2x2-reflector/geometry.py
View file

@ -44,20 +44,15 @@ def find_assembly(assembly_name, wrap_geometry=True):
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.add_surface(surface=min_z, halfspace=+1)
root_cell.add_surface(surface=max_z, halfspace=-1)
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry()
fuel_assembly.root_universe = root_univ
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly
@ -130,12 +125,7 @@ def build_reflector(assembly1_name, assembly2_name):
max_z = openmc.ZPlane(z0=197.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.add_surface(surface=min_z, halfspace=+1)
root_cell.add_surface(surface=max_z, halfspace=-1)
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe for this fuel assembly
root_univ = openmc.Universe(universe_id=0, name='root universe')

8
assembly/build-fresh.py
View file

@ -21,7 +21,7 @@ openmc_geometry.export_to_xml()
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = True if multipole == 'y' else False
multipole = (multipole == 'y')
# Construct uniform initial source distribution over fissionable zones
lower_left = [-10.70864, -10.70864, +192.5]
@ -52,9 +52,9 @@ beavrs.write_openmc_plots()
plot = openmc.Plot()
plot.width = [10.70864*2, 10.70864*2]
plot.origin = [0., 0., 195.]
plot.color = 'mat'
plot.color_by = 'material'
plot.filename = 'assembly'
plot.col_spec = beavrs.plots.colspec_mat
plot.colors = beavrs.plots.colors_mat
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])
@ -106,7 +106,7 @@ mesh.type = 'regular'
mesh.dimension = [17, 17, 1]
mesh.lower_left = lower_left
mesh.width = (np.array(upper_right) - np.array(lower_left))
mesh.width[:2] /= 17
mesh.width[:2] /= mesh.dimension
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate energy-integrated fission rate mesh Tally

11
assembly/geometry.py
View file

@ -44,20 +44,15 @@ def find_assembly(assembly_name, wrap_geometry=True):
min_z = openmc.ZPlane(z0=192.5, boundary_type='reflective')
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.add_surface(surface=min_z, halfspace=+1)
root_cell.add_surface(surface=max_z, halfspace=-1)
root_cell.region = \
+min_x & -max_x & +min_y & -max_y & +min_z & -max_z
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_assembly = openmc.Geometry()
fuel_assembly.root_universe = root_univ
fuel_assembly = openmc.Geometry(root_univ)
return fuel_assembly

6
pin-cell/build-fresh.py
View file

@ -21,7 +21,7 @@ openmc_geometry.export_to_xml()
# Query the user on whether to use multipole cross sections
multipole = input('Use multipole cross sections? (y/n): ').lower()
multipole = True if multipole == 'y' else False
multipole = (multipole == 'y')
# Construct uniform initial source distribution over fissionable zones
lower_left = [-0.62992, -0.62992, -10.0]
@ -52,9 +52,9 @@ beavrs.write_openmc_plots()
plot = openmc.Plot()
plot.width = [1.25984, 1.25984]
plot.origin = [0., 0., np.inf]
plot.color = 'mat'
plot.color_by = 'material'
plot.filename = 'fuel-pin'
plot.col_spec = beavrs.plots.colspec_mat
plot.colors = beavrs.plots.colors_mat
plot.pixels = [1000, 1000]
plot_file = openmc.Plots([plot])

8
pin-cell/geometry.py
View file

@ -52,18 +52,14 @@ def find_pin(pin_name, wrap_geometry=True):
root_cell.fill = fuel_pin
# Add boundaries to the root Cell
root_cell.add_surface(surface=min_x, halfspace=+1)
root_cell.add_surface(surface=max_x, halfspace=-1)
root_cell.add_surface(surface=min_y, halfspace=+1)
root_cell.add_surface(surface=max_y, halfspace=-1)
root_cell.region = +min_x & -max_x & +min_y & -max_y
# Create a root Universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Create a Geometry
fuel_pin = openmc.Geometry()
fuel_pin.root_universe = root_univ
fuel_pin = openmc.Geometry(root_univ)
return fuel_pin

2
smr/build-fresh.py
View file

@ -87,7 +87,7 @@ mesh.type = 'regular'
mesh.dimension = [7*17, 7*17, 100]
mesh.lower_left = lower_left
mesh.width = (np.array(upper_right) - np.array(lower_left))
mesh.width[:2] /= (7*17)
mesh.width[:2] /= mesh.dimension
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate energy-integrated fission rate mesh Tally

8
smr/smr/plots.py
View file

@ -26,7 +26,7 @@ plots = openmc.Plots()
plot = openmc.Plot(name='radial slice ')
plot.basis = 'xy'
plot.color = 'mat'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [25*lattice_pitch/2, 25*lattice_pitch/2.]
plot.filename = 'radial_xy_slice'
@ -37,7 +37,7 @@ plots += [plot]
plot = openmc.Plot(name='axial slice')
plot.basis = 'xz'
plot.color = 'mat'
plot.color_by = 'material'
plot.origin = [0., 0., (highest_extent-lowest_extent)/2.]
plot.width = [rpv_OR*2., (highest_extent-lowest_extent)]
plot.filename = 'axial_xz_slice'
@ -48,7 +48,7 @@ plots += [plot]
plot = openmc.Plot(name='assembly grid spacer')
plot.basis = 'xy'
plot.color = 'mat'
plot.color_by = 'material'
plot.origin = [0., 0., 102.021]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_grid_spacer'
@ -59,7 +59,7 @@ plots += [plot]
plot = openmc.Plot(name='assembly no spacer')
plot.basis = 'xy'
plot.color = 'mat'
plot.color_by = 'material'
plot.origin = [0., 0., 90.]
plot.width = [lattice_pitch*1.5, lattice_pitch*1.5]
plot.filename = 'assm_no_spacer'