<?xml version='1.0' encoding='utf-8'?>
<model>
  <materials>
    <cross_sections>2g.h5</cross_sections>
    <material id="1" name="mat_1">
      <density units="macro" value="1.0"/>
      <macroscopic name="mat_1"/>
    </material>
  </materials>
  <geometry>
    <cell id="1" material="1" region="1 -2" universe="0"/>
    <surface boundary="reflective" coeffs="0.0" id="1" type="x-plane"/>
    <surface boundary="vacuum" coeffs="929.45" id="2" type="x-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="box">
        <parameters>0.0 -1000.0 -1000.0 929.45 1000.0 1000.0</parameters>
      </space>
    </source>
    <output>
      <summary>false</summary>
    </output>
    <energy_mode>multi-group</energy_mode>
    <tabular_legendre>
      <enable>false</enable>
    </tabular_legendre>
  </settings>
  </model>
