OpenMC/tests/regression_tests/mg_temperature_multi/inputs_true.dat

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<?xml version='1.0' encoding='utf-8'?>
<model>
<materials>
<cross_sections>mgxs.h5</cross_sections>
<material id="1" name="UO2 fuel">
<density value="1.0" units="macro"/>
<macroscopic name="UO2"/>
</material>
<material id="2" name="Water">
<density value="1.0" units="macro"/>
<macroscopic name="LWTR"/>
</material>
</materials>
<geometry>
<cell id="1" name="fuel inner" material="1" region="-1" temperature="600.0" universe="1"/>
<cell id="2" name="fuel outer" material="1" region="1 -2" temperature="294.0" universe="1"/>
<cell id="3" name="moderator" material="2" region="2 3 -4 5 -6" universe="1"/>
<surface id="1" name="Fuel IR" type="z-cylinder" coeffs="0.0 0.0 0.25"/>
<surface id="2" name="Fuel OR" type="z-cylinder" coeffs="0.0 0.0 0.54"/>
<surface id="3" name="minimum x" type="x-plane" boundary="reflective" coeffs="-0.63"/>
<surface id="4" name="maximum x" type="x-plane" boundary="reflective" coeffs="0.63"/>
<surface id="5" name="minimum y" type="y-plane" boundary="reflective" coeffs="-0.63"/>
<surface id="6" name="maximum y" type="y-plane" boundary="reflective" coeffs="0.63"/>
</geometry>
<settings>
<run_mode>eigenvalue</run_mode>
<particles>1000</particles>
<batches>10</batches>
<inactive>5</inactive>
<source type="independent" strength="1.0" particle="neutron">
<space type="box">
<parameters>-0.63 -0.63 -1 0.63 0.63 1</parameters>
</space>
<constraints>
<fissionable>true</fissionable>
</constraints>
</source>
<energy_mode>multi-group</energy_mode>
</settings>
<tallies>
<filter id="1" type="cell">
<bins>1</bins>
</filter>
<filter id="2" type="cell">
<bins>2</bins>
</filter>
<tally id="1" name="inner tally">
<filters>1</filters>
<scores>flux</scores>
</tally>
<tally id="2" name="outer tally">
<filters>2</filters>
<scores>flux</scores>
</tally>
</tallies>
</model>