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Fixed inputs in examples directory.
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13 changed files with 41 additions and 41 deletions
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@ -1,17 +1,18 @@
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<?xml version="1.0"?>
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<materials>
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<!-- Definition of materials -->
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<default_xs>70c</default_xs>
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<material id="40">
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<density value="4.5" units="g/cc" />
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<nuclide name="U-235" xs="03c" ao="1.0" />
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<nuclide name="U-235" ao="1.0" />
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</material>
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<material id="41">
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<density value="1.0" units="g/cc" />
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<nuclide name="H-1" xs="03c" ao="2.0" />
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<nuclide name="O-16" xs="03c" ao="1.0" />
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<sab name="lwe7" xs="00t" />
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<nuclide name="H-1" ao="2.0" />
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<nuclide name="O-16" ao="1.0" />
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<sab name="lwtr" xs="10t" />
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</material>
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</materials>
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@ -1,11 +1,6 @@
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<?xml version="1.0" encoding="UTF-8"?>
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<settings>
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<!-- Path for cross section library -->
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<cross_sections>
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/home/paulromano/openmc/cross_sections_serpent.xml
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</cross_sections>
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<!-- Parameters for criticality calculation -->
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<criticality>
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<cycles>15</cycles>
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@ -5,7 +5,7 @@
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<filters>
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<cell>100</cell>
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</filters>
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<macros>total scatter nu-scatter absorption fission nu-fission</macros>
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<scores>total scatter nu-scatter absorption fission nu-fission</scores>
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</tally>
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<tally id="2">
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@ -13,7 +13,7 @@
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<cell>100</cell>
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<energy>0 20.0</energy>
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</filters>
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<macros>total scatter nu-scatter absorption fission nu-fission</macros>
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<scores>total scatter nu-scatter absorption fission nu-fission</scores>
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</tally>
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<tally id="3">
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@ -22,7 +22,7 @@
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<energy>0 20.0</energy>
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<energyout>0 20.0</energyout>
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</filters>
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<macros>scatter nu-scatter nu-fission</macros>
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<scores>scatter nu-scatter nu-fission</scores>
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</tally>
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</tallies>
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