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