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Added sphereoutput compare openmc test
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sbradnam committed Mar 19, 2024
1 parent e3b8132 commit 669d4d2
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Showing 16 changed files with 1,339 additions and 2 deletions.
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<geometry>
<surface boundary="transmission" coeffs="0.0 0.0 0.0 5.0" id="1" type="sphere" />
<surface boundary="vacuum" coeffs="0.0 0.0 0.0 50.0" id="2" type="sphere" />
<surface boundary="transmission" coeffs="0.0 0.0 0.0 60.0" id="3" type="sphere" />
<cell id="1" material="void" region=" -1" universe="0" />
<cell id="2" material="1" region=" 1 -2" universe="0" />
<cell id="3" material="void" region=" 2" universe="0" />
</geometry>
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<materials>
<material id="1" name="m1">
<density units="g/cc" value="0.0708" />
<nuclide ao="1.0" name="H1" />
</material>
</materials>
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<?xml version='1.0' encoding='utf-8'?>
<settings>
<particles>10</particles>
<batches>100</batches>
<run_mode>fixed source</run_mode>
<source strength="1.0">
<space type="point">
<parameters>0 0 0</parameters>
</space>
<energy interpolation="histogram" type="tabular">
<parameters>1.0 100000.0 1000000.0 10000000.0 14000000.0 0.0 1.0 1.0 1.0 1.0</parameters>
</energy>
</source>
<photon_transport>true</photon_transport>
</settings>
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<?xml version='1.0' encoding='utf-8'?>
<tallies>
<filter id="1" type="cell">
<bins>2</bins>
</filter>
<filter id="2" type="particle">
<bins>neutron</bins>
</filter>
<filter id="3" type="particle">
<bins>photon</bins>
</filter>
<filter id="5" type="energy">
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<filter id="6" type="energy">
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</bins>
</filter>
<tally id="4" name="Neutron spectra">
<filters>1 2 5</filters>
<scores>flux</scores>
</tally>
<tally id="14" name="Photon spectra">
<filters>1 3 6</filters>
<scores>flux</scores>
</tally>
</tallies>
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<geometry>
<surface boundary="transmission" coeffs="0.0 0.0 0.0 5.0" id="1" type="sphere" />
<surface boundary="vacuum" coeffs="0.0 0.0 0.0 50.0" id="2" type="sphere" />
<surface boundary="transmission" coeffs="0.0 0.0 0.0 60.0" id="3" type="sphere" />
<cell id="1" material="void" region=" -1" universe="0" />
<cell id="2" material="1" region=" 1 -2" universe="0" />
<cell id="3" material="void" region=" 2" universe="0" />
</geometry>
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<materials>
<material id="1" name="m1">
<density units="g/cc" value="3.18" />
<nuclide name="Ca40" wo="0.496192" />
<nuclide name="Ca42" wo="0.003477076" />
<nuclide name="Ca43" wo="0.000742804" />
<nuclide name="Ca44" wo="0.011744" />
<nuclide name="Ca46" wo="2.354339e-05" />
<nuclide name="Ca48" wo="0.001148528" />
<nuclide name="F19" wo="0.486672" />
</material>
</materials>
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<?xml version='1.0' encoding='utf-8'?>
<settings>
<particles>10</particles>
<batches>100</batches>
<run_mode>fixed source</run_mode>
<source strength="1.0">
<space type="point">
<parameters>0 0 0</parameters>
</space>
<energy interpolation="histogram" type="tabular">
<parameters>1.0 100000.0 1000000.0 10000000.0 14000000.0 0.0 1.0 1.0 1.0 1.0</parameters>
</energy>
</source>
<photon_transport>true</photon_transport>
</settings>
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