<?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="U235"/>
    </material>
  </materials>
  <geometry>
    <cell id="1" material="1" region="-1" universe="1"/>
    <surface boundary="reflective" coeffs="0.0 0.0 0.0 100.0" id="1" type="sphere"/>
  </geometry>
  <settings>
    <run_mode>eigenvalue</run_mode>
    <particles>1000</particles>
    <batches>5</batches>
    <inactive>2</inactive>
    <source particle="neutron" strength="1.0" type="independent">
      <space type="point">
        <parameters>0 0 0</parameters>
      </space>
    </source>
    <photon_transport>true</photon_transport>
  </settings>
  <tallies>
    <filter id="1" type="particle">
      <bins>neutron photon</bins>
    </filter>
    <tally id="1">
      <filters>1</filters>
      <nuclides>U235 total</nuclides>
      <scores>fission heating-local</scores>
      <estimator>tracklength</estimator>
    </tally>
    <tally id="2">
      <filters>1</filters>
      <nuclides>U235 total</nuclides>
      <scores>fission heating heating-local</scores>
      <estimator>collision</estimator>
    </tally>
    <tally id="3">
      <filters>1</filters>
      <nuclides>U235 total</nuclides>
      <scores>fission heating heating-local</scores>
      <estimator>analog</estimator>
    </tally>
  </tallies>
</model>
