crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-29 06:35:48 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

merged with upstream/develop

Browse source
This commit is contained in:
Sam Shaner 2016-11-03 16:33:17 -04:00
commit 1e90230536
109 changed files with 1857 additions and 1957 deletions
Download patch file Download diff file Expand all files Collapse all files

6
examples/xml/basic/tallies.xml
View file

@ -8,14 +8,14 @@
<tally id="2">
<filter type="cell" bins="100" />
<filter type="energy" bins="0 20.0" />
<filter type="energy" bins="0 20.0e6" />
<scores>total scatter nu-scatter absorption fission nu-fission</scores>
</tally>
<tally id="3">
<filter type="cell" bins="100" />
<filter type="energy" bins="0 20.0" />
<filter type="energyout" bins="0 20.0" />
<filter type="energy" bins="0 20.0e6" />
<filter type="energyout" bins="0 20.0e6" />
<scores>scatter nu-scatter nu-fission</scores>
</tally>

2
examples/xml/pincell/tallies.xml
View file

@ -9,7 +9,7 @@
<tally id="1">
<filter type="mesh" bins="1" />
<filter type="energy" bins="0. 4.e-6 20.0" />
<filter type="energy" bins="0. 4. 20.0e6" />
<scores>flux fission nu-fission</scores>
</tally>

9
examples/xml/pincell_multigroup/geometry.xml
View file

@ -1,11 +1,10 @@
<?xml version='1.0' encoding='utf-8'?>
<geometry>
<cell id="1" material="1" name="cell 1" region="-1" universe="0" />
<cell id="2" material="7" name="cell 2" region="((((1 4) -5) 6) -7)" 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="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" />
<surface boundary="reflective" coeffs="0.63" id="7" name="top" type="y-plane" />
</geometry>

62
examples/xml/pincell_multigroup/materials.xml
View file

@ -1,54 +1,12 @@
<?xml version="1.0"?>
<?xml version='1.0' encoding='utf-8'?>
<materials>
<cross_sections>./mgxs.h5</cross_sections>
<!-- 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>
<cross_sections>./mgxs.h5</cross_sections>
<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
View file

Binary file not shown.

47
examples/xml/pincell_multigroup/settings.xml
View file

@ -1,37 +1,14 @@
<?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>
<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>
<energy_mode>multi-group</energy_mode>
</settings>

2
examples/xml/pincell_multigroup/tallies.xml
View file

@ -6,7 +6,7 @@
<upper_right>0.63 0.63 1e+50</upper_right>
</mesh>
<tally id="1" name="tally 1">
<filter bins="1e-11 6.35e-08 1e-05 0.0001 0.001 0.5 1.0 20.0" type="energy" />
<filter bins="1e-05 0.0635 10.0 100.0 1000.0 500000.0 1000000.0 20000000.0" type="energy" />
<filter bins="1" type="mesh" />
<scores>flux fission nu-fission</scores>
</tally>