<?xml version='1.0' encoding='utf-8'?>
<model>
  <materials>
    <material depletable="true" id="1" name="mat">
      <density units="atom/b-cm" value="0.069335"/>
      <nuclide ao="40.0" name="H1"/>
      <nuclide ao="1.0" name="U235"/>
    </material>
  </materials>
  <geometry>
    <cell id="1" material="1" name="box" region="1 -2 3 -4 5 -6" universe="0"/>
    <surface boundary="reflective" coeffs="-1" id="1" type="x-plane"/>
    <surface boundary="reflective" coeffs="1" id="2" type="x-plane"/>
    <surface boundary="reflective" coeffs="-1" id="3" type="y-plane"/>
    <surface boundary="reflective" coeffs="1" id="4" type="y-plane"/>
    <surface boundary="reflective" coeffs="-1" id="5" type="z-plane"/>
    <surface boundary="reflective" coeffs="1" id="6" type="z-plane"/>
  </geometry>
  <settings>
    <run_mode>fixed source</run_mode>
    <particles>100</particles>
    <batches>10</batches>
    <source particle="neutron" strength="1.0" type="independent">
      <space type="box">
        <parameters>-1 -1 -1 1 1 1</parameters>
      </space>
      <energy parameters="988000.0 2.249e-06" type="watt"/>
    </source>
    <create_fission_neutrons>false</create_fission_neutrons>
  </settings>
  <tallies>
    <tally id="1">
      <scores>flux</scores>
    </tally>
  </tallies>
</model>
