<?xml version='1.0' encoding='utf-8'?>
<model>
  <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>
  <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>
  <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>
