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corrected syntax in PyAPI test, and deleted forgotten cellto options in input handling

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guillaume 2017-08-06 13:54:03 -04:00
parent 5b2aae3b05
commit f0c051db5f
3 changed files with 84 additions and 66 deletions
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3
openmc/filter.py
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@ -18,8 +18,7 @@ AUTO_FILTER_ID = 10000
_FILTER_TYPES = ['universe', 'material', 'cell', 'cellborn', 'surface',
'mesh', 'energy', 'energyout', 'mu', 'polar', 'azimuthal',
'distribcell', 'delayedgroup', 'energyfunction', 'cellfrom',
'cellto']
'distribcell', 'delayedgroup', 'energyfunction', 'cellfrom']
_CURRENT_NAMES = {1: 'x-min out', 2: 'x-min in',
3: 'x-max out', 4: 'x-max in',

2
src/input_xml.F90
View file

@ -3004,7 +3004,7 @@ contains
end if
n_words = node_word_count(node_filt, "bins")
case ("mesh", "universe", "material", "cell", "distribcell", &
"cellborn", "cellto", "cellfrom", "surface", "delayedgroup")
"cellborn", "cellfrom", "surface", "delayedgroup")
if (.not. check_for_node(node_filt, "bins")) then
call fatal_error("Bins not set in filter " // trim(to_str(filter_id)))
end if

145
tests/test_surface_tally/test_surface_tally.py
View file

@ -8,7 +8,8 @@ import numpy as np
import openmc
import pandas as pd
class CreateSurfaceTallyTestHarness(PyAPITestHarness):
class SurfaceTallyTestHarness(PyAPITestHarness):
def _build_inputs(self):
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(name='UO2 fuel at 2.4% wt enrichment')
@ -25,15 +26,15 @@ class CreateSurfaceTallyTestHarness(PyAPITestHarness):
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, borated_water])
materials_file.export_to_xml()
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.4, name='Fuel OR')
left = openmc.XPlane(surface_id=2, x0=-0.62992, name='left')
right = openmc.XPlane(surface_id=3, x0=0.62992, name='right')
bottom = openmc.YPlane(y0=-0.62992, name='bottom')
top = openmc.YPlane(y0=0.62992, name='top')
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=1, \
name='Fuel OR')
left = openmc.XPlane(surface_id=2, x0=-2, name='left')
right = openmc.XPlane(surface_id=3, x0=2, name='right')
bottom = openmc.YPlane(y0=-2, name='bottom')
top = openmc.YPlane(y0=2, name='top')
left.boundary_type = 'vacuum'
right.boundary_type = 'reflective'
@ -49,7 +50,7 @@ class CreateSurfaceTallyTestHarness(PyAPITestHarness):
water.region = +fuel_or & -right & +bottom & -top
# Register Materials with Cells
fuel.fill = uo2
fuel.fill = uo2
water.fill = borated_water
# Instantiate pin cell Universe
@ -63,82 +64,101 @@ class CreateSurfaceTallyTestHarness(PyAPITestHarness):
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Instantiate a Geometry, register the root Universe, and export to XML
# Instantiate a Geometry, register the root Universe
geometry = openmc.Geometry(root_univ)
geometry.export_to_xml()
# Instantiate a Settings object, set all runtime parameters, and export to XML
# Instantiate a Settings object, set all runtime parameters
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 0
settings_file.particles = 1000
#settings_file.output = {'tallies': True}
# Create an initial uniform spatial source distribution over fissionable zones
# Create an initial uniform spatial source distribution
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:],\
only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.export_to_xml()
# Tallies file
tallies_file = openmc.Tallies()
# Cell to cell tallies
# These filters are same for all tallies
# Create partial current tallies from fuel to water
# Filters
two_groups = np.array([0., 4, 20.]) * 1e6
energy_filter = openmc.EnergyFilter(two_groups)
polar_filter = openmc.PolarFilter([0, np.pi / 4, np.pi])
azimuthal_filter = openmc.AzimuthalFilter([0, np.pi / 4, np.pi])
cell_to_cell_tallies = []
tally_index = 0
for cell1 in pin_cell.get_all_cells():
for cell2 in pin_cell.get_all_cells():
if cell1 != cell2 and abs(abs(cell1-cell2)-1) < 0.1: # no need for cell1 to cell1
# Cell to cell filters for partial current
cell_from_filter = openmc.CellFromFilter(cell1)
cell_to_filter = openmc.CellFilter(cell2)
cell_to_cell_tallies.append(openmc.Tally(tally_id=tally_index, name=str(cell1)+'-'+str(cell2)))
cell_to_cell_tallies[tally_index].filters = [cell_from_filter, cell_to_filter, energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tallies[tally_index].scores = ['current']
tallies_file.append(cell_to_cell_tallies[tally_index])
tally_index += 1
# Cell from + surface filters for partial current
surface_filter = openmc.SurfaceFilter([1])
cell_to_cell_tallies.append(openmc.Tally(tally_id=tally_index, name=str(cell1)+'-surface1'))
cell_to_cell_tallies[tally_index].filters = [cell_from_filter, surface_filter, energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tallies[tally_index].scores = ['current']
tallies_file.append(cell_to_cell_tallies[tally_index])
tally_index += 1
# Surface filter on inner surface, for net current
energy_filter = openmc.EnergyFilter(two_groups)
polar_filter = openmc.PolarFilter([0, np.pi / 4, np.pi])
azimuthal_filter = openmc.AzimuthalFilter([0, np.pi / 4, np.pi])
surface_filter = openmc.SurfaceFilter([1])
surf_tally1 = openmc.Tally(tally_id=tally_index, name='net_cylinder')
surf_tally1.filters = [surface_filter, energy_filter, polar_filter, azimuthal_filter]
surf_tally1.scores = ['current']
cell_from_filter = openmc.CellFromFilter(fuel)
cell_filter = openmc.CellFilter(water)
# Use Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Use a Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create partial current tallies from water to fuel
# Filters
cell_from_filter = openmc.CellFromFilter(water)
cell_filter = openmc.CellFilter(fuel)
# Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create a net current tally on inner surface using a surface filter
surface_filter = openmc.SurfaceFilter([1])
surf_tally1 = openmc.Tally(name='net_cylinder')
surf_tally1.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally1.scores = ['current']
tallies_file.append(surf_tally1)
tally_index += 1
# Surface filter on left surface, vacuum BC, for net current = leakage
# Create a net current tally on left surface using a surface filter
# This surface has a vacuum boundary condition, so leakage is tallied
surface_filter = openmc.SurfaceFilter([2])
surf_tally2 = openmc.Tally(tally_id=tally_index, name='leakage_left')
surf_tally2.filters = [surface_filter, energy_filter, polar_filter, azimuthal_filter]
surf_tally2.scores = ['current']
surf_tally2 = openmc.Tally(name='leakage_left')
surf_tally2.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally2.scores = ['current']
tallies_file.append(surf_tally2)
tally_index += 1
# Surface filter on right surface, reflective, current tally doesnt pick up the zero net current though
# Create a net current tally on right surface using a surface filter
# This surface has a reflective boundary condition, but the zero
# net current is not picked up because particles are only tallied once
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(tally_id=tally_index, name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
tally_index += 1
tallies_file.export_to_xml()
@ -153,12 +173,11 @@ class CreateSurfaceTallyTestHarness(PyAPITestHarness):
df = df.append(t.get_pandas_dataframe(), ignore_index=True)
# Extract the relevant data as a CSV string.
cols = ('d_material', 'd_nuclide', 'd_variable', 'score', 'mean',
'std. dev.')
cols = ('mean', 'std. dev.')
return df.to_csv(None, columns=cols, index=False, float_format='%.7e')
return outstr
if __name__ == '__main__':
harness = CreateSurfaceTallyTestHarness('statepoint.10.h5')
harness = SurfaceTallyTestHarness('statepoint.10.h5')
harness.main()