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Use python API for setup in test_deplete_fission_yields

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Andrew Johnson 2019-08-30 09:55:58 -05:00
parent 399b77e38f
commit c486b4187f
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1 changed files with 24 additions and 44 deletions
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68
tests/unit_tests/test_deplete_fission_yields.py
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@ -6,7 +6,8 @@ from unittest.mock import Mock
import bisect
import pytest
import numpy
import numpy as np
import openmc
from openmc import capi
from openmc.deplete.nuclide import Nuclide, FissionYieldDistribution
from openmc.deplete.helpers import (
@ -19,46 +20,25 @@ def materials(tmpdir_factory):
"""Use C API to construct realistic materials for testing tallies"""
tmpdir = tmpdir_factory.mktemp("capi")
orig = tmpdir.chdir()
# Create proxy xml files to please openmc
with open("geometry.xml", "w") as stream:
stream.write("""
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="1" material="1" name="fuel" region="1 -2 3 -4" universe="1" />
<surface boundary="reflective" coeffs="-0.63" id="1" type="x-plane" />
<surface boundary="reflective" coeffs="0.63" id="2" type="x-plane" />
<surface boundary="reflective" coeffs="-0.63" id="3" type="y-plane" />
<surface boundary="reflective" coeffs="0.63" id="4" type="y-plane" />
</geometry>
""")
with open("settings.xml", "w") as stream:
stream.write("""
<?xml version='1.0' encoding='utf-8'?>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>100</particles>
<batches>10</batches>
<inactive>0</inactive>
<verbosity>1</verbosity>
<source strength="1.0">
<space type="point">
<parameters>0.0 0.0 0.0</parameters>
</space>
</source>
</settings>
""")
with open("materials.xml", "w") as stream:
stream.write("""
<?xml version='1.0' encoding='utf-8'?>
<materials>
<material depletable="true" id="1" name="U" volume="71.67537585">
<density units="g/cc" value="10.4" />
<nuclide ao="1.0" name="U235" />
<nuclide ao="1.0" name="U238" />
<nuclide ao="1.0" name="Xe135" />
<nuclide ao="1.0" name="Pu239" />
</material>
</materials>""")
# Create proxy problem to please openmc
mfuel = openmc.Material(name="test_fuel")
mfuel.volume = 1.0
for nuclide in ["U235", "U238", "Xe135", "Pu239"]:
mfuel.add_nuclide(nuclide, 1.0)
openmc.Materials([mfuel]).export_to_xml()
# Geometry
box = openmc.rectangular_prism(1.0, 1.0, boundary_type="reflective")
cell = openmc.Cell(fill=mfuel, region=box)
root = openmc.Universe(cells=[cell])
openmc.Geometry(root).export_to_xml()
# settings
settings = openmc.Settings()
settings.particles = 100
settings.inactive = 0
settings.batches = 10
settings.verbosity = 1
settings.export_to_xml()
try:
with capi.run_in_memory():
yield [capi.Material(), capi.Material()]
@ -87,7 +67,7 @@ def proxy_tally_data(tally, fill=None):
if isinstance(tfilter, capi.EnergyFilter):
this_bins -= 1
n_bins *= max(this_bins, 1)
data = numpy.empty((n_bins, n_nucs * n_scores, 3))
data = np.empty((n_bins, n_nucs * n_scores, 3))
if fill is not None:
data.fill(fill)
return data
@ -229,7 +209,7 @@ def test_cutoff_helper(materials, nuclide_bundle, therm_frac):
helper.unpack()
# expected results of shape (n_mats, 2, n_tnucs)
expected_results = numpy.empty((1, 2, len(tally_nucs)))
expected_results = np.empty((1, 2, len(tally_nucs)))
expected_results[:, 0] = therm_frac
expected_results[:, 1] = 1 - therm_frac
assert helper.results == pytest.approx(expected_results)
@ -285,7 +265,7 @@ def test_averaged_helper(materials, nuclide_bundle, avg_energy):
helper._weighted_tally.results = weighted_results
helper.unpack()
expected_results = numpy.ones((1, len(tallied_nucs))) * avg_energy
expected_results = np.ones((1, len(tallied_nucs))) * avg_energy
assert helper.results == pytest.approx(expected_results)
actual_yields = helper.weighted_yields(0)