OpenMC/tests/regression_tests/tally_arithmetic/inputs_true.dat

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<?xml version='1.0' encoding='utf-8'?>
<model>
<materials>
<material id="1" depletable="true">
<density value="10.0" units="g/cm3"/>
<nuclide name="U234" ao="1.0"/>
<nuclide name="U235" ao="4.0"/>
<nuclide name="U238" ao="95.0"/>
</material>
<material id="2" name="light water">
<density value="1.0" units="g/cm3"/>
<nuclide name="H1" ao="2.0"/>
<nuclide name="O16" ao="1.0"/>
<sab name="c_H_in_H2O"/>
</material>
</materials>
<geometry>
<cell id="1" material="1" region="-1" universe="1"/>
<cell id="2" material="2" region="1 -2" universe="1"/>
<surface id="1" type="z-cylinder" coeffs="0.0 0.0 5.0"/>
<surface id="2" type="z-cylinder" boundary="vacuum" coeffs="0.0 0.0 10.0"/>
</geometry>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>1000</particles>
<batches>5</batches>
<inactive>0</inactive>
</settings>
<tallies>
<mesh id="1">
<dimension>2 2</dimension>
<lower_left>-10.0 -10.0</lower_left>
<upper_right>10.0 10.0</upper_right>
</mesh>
<filter id="2" type="material">
<bins>1 2</bins>
</filter>
<filter id="1" type="energy">
<bins>0.0 10.0 20000000.0</bins>
</filter>
<filter id="3" type="mesh">
<bins>1</bins>
</filter>
<tally id="1" name="tally 1">
<filters>2 1</filters>
<nuclides>U234 U235</nuclides>
<scores>nu-fission total</scores>
</tally>
<tally id="2" name="tally 2">
<filters>1 3</filters>
<nuclides>U238 U235</nuclides>
<scores>total fission</scores>
</tally>
</tallies>
</model>