<?xml version='1.0' encoding='utf-8'?>
<geometry>
  <cell id="1" material="1" region="-1" universe="1" />
  <cell id="2" material="2" region="1 -2" universe="1" />
  <surface coeffs="0.0 0.0 5.0" id="1" type="z-cylinder" />
  <surface boundary="vacuum" coeffs="0.0 0.0 10.0" id="2" type="z-cylinder" />
</geometry>
<?xml version='1.0' encoding='utf-8'?>
<materials>
  <material depletable="true" id="1">
    <density units="g/cm3" value="10.0" />
    <nuclide ao="1.0" name="U234" />
    <nuclide ao="4.0" name="U235" />
    <nuclide ao="95.0" name="U238" />
  </material>
  <material id="2" name="light water">
    <density units="g/cm3" value="1.0" />
    <nuclide ao="2.0" name="H1" />
    <nuclide ao="1.0" name="O16" />
    <sab name="c_H_in_H2O" />
  </material>
</materials>
<?xml version='1.0' encoding='utf-8'?>
<settings>
  <run_mode>eigenvalue</run_mode>
  <particles>1000</particles>
  <batches>5</batches>
  <inactive>0</inactive>
</settings>
<?xml version='1.0' encoding='utf-8'?>
<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>
