OpenMC/tests/regression_tests/ifp/total/inputs_true.dat
Alex Nellis 781dbf9c12
Multi-group capability for kinetics parameter calculations with Iterated Fission Probability (#3425)
Co-authored-by: GuySten <guyste@post.bgu.ac.il>
Co-authored-by: Paul Romano <paul.k.romano@gmail.com>
2025-09-26 16:27:05 +00:00

33 lines
1,009 B
XML

<?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>