OpenMC/tests/regression_tests/ifp/total/inputs_true.dat

<?xml version='1.0' encoding='utf-8'?>
<model>
  <materials>
    <material id="1" name="core" depletable="true">
      <density value="16.0" units="g/cm3"/>
      <nuclide name="U235" ao="1.0"/>
    </material>
  </materials>
  <geometry>
    <cell id="1" material="1" region="-1" universe="1"/>
    <surface id="1" type="sphere" boundary="vacuum" coeffs="0.0 0.0 0.0 10.0"/>
  </geometry>
  <settings>
    <run_mode>eigenvalue</run_mode>
    <particles>1000</particles>
    <batches>20</batches>
    <inactive>5</inactive>
    <source type="independent" strength="1.0" particle="neutron">
      <space type="box">
        <parameters>-10.0 -10.0 -10.0 10.0 10.0 10.0</parameters>
      </space>
      <constraints>
        <fissionable>true</fissionable>
      </constraints>
    </source>
    <ifp_n_generation>5</ifp_n_generation>
  </settings>
  <tallies>
    <tally id="1" name="ifp-scores">
      <scores>ifp-time-numerator ifp-beta-numerator ifp-denominator</scores>
    </tally>
  </tallies>
</model>