OpenMC/tests/regression_tests/fixed_source/inputs_true.dat

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XML

<?xml version='1.0' encoding='utf-8'?>
<model>
<materials>
<material id="1" depletable="true">
<density value="7.5" units="g/cc"/>
<nuclide name="O16" ao="1.0"/>
<nuclide name="U238" ao="0.0001"/>
</material>
</materials>
<geometry>
<cell id="1" material="1" region="-1" universe="1"/>
<surface id="1" type="sphere" boundary="vacuum" coeffs="0.0 0.0 0.0 10.0"/>
</geometry>
<settings>
<run_mode>fixed source</run_mode>
<particles>100</particles>
<batches>10</batches>
<source type="independent" strength="10.0" particle="neutron">
<space type="point">
<parameters>0.0 0.0 0.0</parameters>
</space>
</source>
<temperature_default>294</temperature_default>
</settings>
<tallies>
<tally id="1">
<scores>flux</scores>
</tally>
</tallies>
</model>