<?xml version='1.0' encoding='utf-8'?>
<model>
  <materials>
    <material id="1">
      <density units="g/cc" value="1.0"/>
      <nuclide ao="2.0" name="H1"/>
      <nuclide ao="1.0" name="O16"/>
      <sab name="c_H_in_H2O"/>
    </material>
    <material depletable="true" id="2">
      <density units="g/cc" value="4.5"/>
      <nuclide ao="1.0" name="U235"/>
    </material>
  </materials>
  <geometry>
    <cell id="1" material="1" universe="1"/>
    <cell id="11" material="2" region="-1" temperature="500 700 0 800" universe="11"/>
    <cell id="12" material="1" region="1" universe="11"/>
    <cell fill="101" id="101" region="2 -3 4 -5" universe="0"/>
    <lattice id="101">
      <pitch>2.0 2.0</pitch>
      <outer>1</outer>
      <dimension>2 2</dimension>
      <lower_left>-2.0 -2.0</lower_left>
      <universes>
11 11 
11 11 </universes>
    </lattice>
    <surface coeffs="0.0 0.0 0.3" id="1" type="z-cylinder"/>
    <surface boundary="reflective" coeffs="-3.0" id="2" type="x-plane"/>
    <surface boundary="reflective" coeffs="3.0" id="3" type="x-plane"/>
    <surface boundary="reflective" coeffs="-3.0" id="4" type="y-plane"/>
    <surface boundary="reflective" coeffs="3.0" id="5" type="y-plane"/>
  </geometry>
  <settings>
    <run_mode>eigenvalue</run_mode>
    <particles>1000</particles>
    <batches>5</batches>
    <inactive>0</inactive>
    <source particle="neutron" strength="1.0" type="independent">
      <space type="box">
        <parameters>-1 -1 -1 1 1 1</parameters>
      </space>
    </source>
    <temperature_multipole>true</temperature_multipole>
    <temperature_tolerance>1000</temperature_tolerance>
  </settings>
  <tallies>
    <tally id="1">
      <nuclides>U235 O16 total</nuclides>
      <scores>total fission (n,gamma) elastic (n,p)</scores>
    </tally>
  </tallies>
</model>
