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fixed issue in mgxs_header.F90 where nu cannot be calculated if a fission xs has not been given

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This commit is contained in:
Sam Shaner 2016-10-13 16:54:47 -04:00
parent 89149ea7eb
commit 480d2ca6e5
9 changed files with 74 additions and 204 deletions
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22
examples/xml/pincell_multigroup/geometry.xml
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@ -1,16 +1,10 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<surface coeffs="0. 0. 0.540" id="1" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="20" type="x-plane" />
<surface boundary="reflective" coeffs=" 0.63" id="21" type="x-plane" />
<surface boundary="reflective" coeffs="-0.63" id="22" type="y-plane" />
<surface boundary="reflective" coeffs=" 0.63" id="23" type="y-plane" />
<cell id="1" material="1" region=" -1" />
<cell id="2" material="7" region="1 20 -21 22 -23" />
<cell id="1" material="1" name="cell 1" region="-1" universe="0" />
<cell id="2" material="2" name="cell 2" region="((((1 4) -5) 6) -7)" universe="0" />
<surface coeffs="0 0 0.54" id="1" name="Fuel OR" type="z-cylinder" />
<surface boundary="reflective" coeffs="-0.63" id="4" name="left" type="x-plane" />
<surface boundary="reflective" coeffs="0.63" id="5" name="right" type="x-plane" />
<surface boundary="reflective" coeffs="-0.63" id="6" name="bottom" type="y-plane" />
<surface boundary="reflective" coeffs="0.63" id="7" name="top" type="y-plane" />
</geometry>

58
examples/xml/pincell_multigroup/materials.xml
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@ -1,51 +1,11 @@
<?xml version="1.0"?>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<!-- UO2 -->
<material id="1">
<density units="macro" value="1.0" />
<macroscopic name="UO2"/>
</material>
<!-- 4.3% MOX -->
<material id="2">
<density units="macro" value="1.0" />
<macroscopic name="MOX1"/>
</material>
<!-- 7.0 MOX -->
<material id="3">
<density units="macro" value="1.0" />
<macroscopic name="MOX2"/>
</material>
<!-- 8.0% MOX -->
<material id="4">
<density units="macro" value="1.0" />
<macroscopic name="MOX3"/>
</material>
<!-- Fission Chamber -->
<material id="5">
<density units="macro" value="1.0" />
<macroscopic name="FC"/>
</material>
<!-- Guide Tube -->
<material id="6">
<density units="macro" value="1.0" />
<macroscopic name="GT"/>
</material>
<!-- Water -->
<material id="7">
<density units="macro" value="1.0" />
<macroscopic name="LWTR"/>
</material>
<!-- Control Rod -->
<material id="8">
<density units="macro" value="1.0" />
<macroscopic name="CR"/>
</material>
<material id="1" name="UO2 fuel">
<density units="macro" value="1.0" />
<macroscopic name="UO2" />
</material>
<material id="2" name="Water">
<density units="macro" value="1.0" />
<macroscopic name="LWTR" />
</material>
</materials>

BIN
examples/xml/pincell_multigroup/mgxs.h5
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50
examples/xml/pincell_multigroup/settings.xml
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@ -1,39 +1,15 @@
<?xml version="1.0"?>
<?xml version='1.0' encoding='utf-8'?>
<settings>
<energy_mode>multi-group</energy_mode>
<!--
Define how many particles to run and for how many batches
in an eigenvalue calculation mode
-->
<eigenvalue>
<batches>100</batches>
<inactive>10</inactive>
<particles>1000</particles>
</eigenvalue>
<!--
Start with uniformally distributed neutron source
with the default energy spectrum of a Maxwellian
and isotropic distribution.
-->
<source>
<space type="box">
<parameters>
-0.63 -0.63 -1E50
0.63 0.63 1E50
</parameters>
</space>
</source>
<output>
<summary>true</summary>
<tallies>true</tallies>
</output>
<survival_biasing>false</survival_biasing>
<cross_sections>./mgxs.h5</cross_sections>
<eigenvalue>
<particles>1000</particles>
<batches>100</batches>
<inactive>10</inactive>
</eigenvalue>
<source strength="1.0">
<space type="box">
<parameters>-0.63 -0.63 -1 0.63 0.63 1</parameters>
</space>
</source>
<cross_sections>./mgxs.h5</cross_sections>
<energy_mode>multi-group</energy_mode>
</settings>

2
examples/xml/pincell_multigroup/tallies.xml
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@ -1,4 +1,4 @@
<?xml version="1.0"?>
<?xml version='1.0' encoding='utf-8'?>
<tallies>
<mesh id="1" type="regular">
<dimension>100 100 1</dimension>