<?xml version='1.0' encoding='utf-8'?>
<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>
<?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="U235" />
  </material>
</materials>
<?xml version='1.0' encoding='utf-8'?>
<settings>
  <run_mode>eigenvalue</run_mode>
  <particles>1000</particles>
  <batches>5</batches>
  <inactive>2</inactive>
  <source strength="1.0">
    <space type="point">
      <parameters>0 0 0</parameters>
    </space>
  </source>
  <photon_transport>true</photon_transport>
</settings>
<?xml version='1.0' encoding='utf-8'?>
<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>
