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53 lines
1.9 KiB
XML
53 lines
1.9 KiB
XML
<?xml version='1.0' encoding='utf-8'?>
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<model>
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<materials>
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<cross_sections>mgxs.h5</cross_sections>
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<material id="1" name="UO2 fuel">
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<density units="macro" value="1.0"/>
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<macroscopic name="UO2"/>
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</material>
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<material id="2" name="Water">
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<density units="macro" value="1.0"/>
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<macroscopic name="LWTR"/>
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</material>
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</materials>
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<geometry>
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<cell id="1" material="1" name="fuel inner" region="-1" temperature="600.0" universe="1"/>
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<cell id="2" material="1" name="fuel outer" region="1 -2" temperature="294.0" universe="1"/>
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<cell id="3" material="2" name="moderator" region="2 3 -4 5 -6" universe="1"/>
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<surface coeffs="0.0 0.0 0.25" id="1" name="Fuel IR" type="z-cylinder"/>
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<surface coeffs="0.0 0.0 0.54" id="2" name="Fuel OR" type="z-cylinder"/>
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<surface boundary="reflective" coeffs="-0.63" id="3" name="minimum x" type="x-plane"/>
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<surface boundary="reflective" coeffs="0.63" id="4" name="maximum x" type="x-plane"/>
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<surface boundary="reflective" coeffs="-0.63" id="5" name="minimum y" type="y-plane"/>
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<surface boundary="reflective" coeffs="0.63" id="6" name="maximum y" type="y-plane"/>
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</geometry>
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<settings>
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<run_mode>eigenvalue</run_mode>
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<particles>1000</particles>
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<batches>10</batches>
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<inactive>5</inactive>
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<source particle="neutron" strength="1.0" type="independent">
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<space type="fission">
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<parameters>-0.63 -0.63 -1 0.63 0.63 1</parameters>
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</space>
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</source>
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<energy_mode>multi-group</energy_mode>
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</settings>
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<tallies>
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<filter id="1" type="cell">
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<bins>1</bins>
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</filter>
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<filter id="2" type="cell">
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<bins>2</bins>
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</filter>
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<tally id="1" name="inner tally">
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<filters>1</filters>
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<scores>flux</scores>
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</tally>
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<tally id="2" name="outer tally">
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<filters>2</filters>
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<scores>flux</scores>
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</tally>
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</tallies>
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</model>
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