OpenMC/tests/test_sourcepoint_interval/materials.xml
2014-01-06 09:57:45 -05:00

9 lines
171 B
XML

<?xml version="1.0"?>
<materials>
<material id="1">
<density value="4.5" units="g/cc" />
<nuclide name="U-235" xs="70c" ao="1.0" />
</material>
</materials>