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added tallies.xml file to tally arithmetic test

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This commit is contained in:
Sam Shaner 2015-12-08 13:27:45 -05:00
parent b17a8e07e3
commit b81fa69db3
2 changed files with 15 additions and 32 deletions
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15
tests/test_tally_arithmetic/tallies.xml Normal file
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@ -0,0 +1,15 @@
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<tally id="10000" name="tally 1">
<filter bins="1" type="cell" />
<filter bins="0.0 20.0" type="energy" />
<nuclides>U-235 U-238</nuclides>
<scores>fission nu-fission</scores>
</tally>
<tally id="10001" name="tally 2">
<filter bins="1" type="material" />
<filter bins="0.0 20.0" type="energy" />
<nuclides>U-235 Pu-239</nuclides>
<scores>fission absorption</scores>
</tally>
</tallies>

32
tests/test_tally_arithmetic/test_tally_arithmetic.py
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@ -10,38 +10,6 @@ import openmc
class TallyArithmeticTestHarness(TestHarness):
def _build_inputs(self):
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
pu239 = openmc.Nuclide('Pu-239')
# Instantiate energy filter
energy_filter = openmc.Filter(type='energy', bins=[0., 20.])
# Create tallies
tally_1 = openmc.Tally(name='tally 1')
tally_1.add_filter(openmc.Filter(type='cell', bins=[1]))
tally_1.add_filter(energy_filter)
tally_1.add_score('fission')
tally_1.add_score('nu-fission')
tally_1.add_nuclide(u235)
tally_1.add_nuclide(u238)
tally_2 = openmc.Tally(name='tally 2')
tally_2.add_filter(openmc.Filter(type='material', bins=[1]))
tally_2.add_filter(energy_filter)
tally_2.add_score('fission')
tally_2.add_score('absorption')
tally_2.add_nuclide(u235)
tally_2.add_nuclide(pu239)
# Export tallies to file
tallies_file = openmc.TalliesFile()
tallies_file.add_tally(tally_1)
tallies_file.add_tally(tally_2)
tallies_file.export_to_xml()
def _get_results(self, hash_output=False):
"""Digest info in the statepoint and return as a string."""