Merge branch 'main' of gitlab:aparler/openmc-designs into main

This commit is contained in:
Adam Parler 2023-10-19 00:10:00 -07:00
commit 2408a4eb88
103 changed files with 21687 additions and 0 deletions

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<?xml version="1.0"?>
<geometry>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<?xml version="1.0"?>
<materials>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<default_xs>71c</default_xs>
<!-- Fuel composed of U-238 O-16 and U-235 -->
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<nuclide name="U-235" xs="71c" wo="0.0185105302348" />
<nuclide name="U-238" xs="71c" wo="0.862943290471" />
<nuclide name="O-16" xs="71c" wo="0.118546179294" />
</material>
<!-- Helium for gap -->
<material id="2">
<density value="1.00000e-4" units="atom/b-cm" />
<nuclide name="He-4" xs="71c" ao="1.0" />
</material>
<!-- Zircaloy 4 -->
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<density value="6.55" units="g/cc" />
<nuclide name="O-16" xs="71c" wo="0.00125" />
<nuclide name="Nb-93" xs="71c" wo="0.01" />
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<?xml version="1.0"?>
<plots>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<plot id="1" color="mat" >
<origin>0. 0. 0.</origin>
<width>25.50 25.50 </width>
<pixels>1000 1000 </pixels>
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</plot>
</plots>

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<?xml version="1.0"?>
<settings>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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/home/ahmedk/Desktop/NuclearCodes/Cross_section_data/nndc/293.6K/cross_sections_mcnp6.1.xml</cross_sections>
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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 48.1251 +/- 1.04779E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 1.23740 +/- 6.41088E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 8.78552E-03 +/- 4.24096E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.503212 +/- 2.62902E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00134 +/- 4.25218E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 33.8552 +/- 9.30306E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 34.8566 +/- 9.65666E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.53152E-04 +/- 1.33191E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.39554E-03 +/- 6.98258E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.37422E-03 +/- 6.79073E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 4.02822E-03 +/- 1.94805E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.30113E-03 +/- 5.88791E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.33259E-04 +/- 2.02954E-07

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<?xml version="1.0"?>
<tallies>
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AP1000/PWR 2.1/summary.h5 Normal file

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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 48.1251 +/- 1.04779E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 1.23740 +/- 6.41088E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 8.78552E-03 +/- 4.24096E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.503212 +/- 2.62902E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00134 +/- 4.25218E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 33.8552 +/- 9.30306E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 34.8566 +/- 9.65666E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.53152E-04 +/- 1.33191E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.39554E-03 +/- 6.98258E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.37422E-03 +/- 6.79073E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 4.02822E-03 +/- 1.94805E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.30113E-03 +/- 5.88791E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.33259E-04 +/- 2.02954E-07

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<?xml version="1.0"?>
<geometry>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<surface id="15" type="y-plane" coeffs="0.627" />
</geometry>

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<?xml version="1.0"?>
<materials>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<default_xs>71c</default_xs>
<!-- Fuel composed of U-238 O-16 and U-235 -->
<material id="1">
<density value="10.5" units="g/cc" />
<nuclide name="U-235" xs="71c" wo="0.0229176" />
<nuclide name="U-238" xs="71c" wo="0.858529" />
<nuclide name="O-16" xs="71c" wo="0.11855276" />
</material>
<!-- Helium for gap -->
<material id="2">
<density value="1.00000e-4" units="atom/b-cm" />
<nuclide name="He-4" xs="71c" ao="1.0" />
</material>
<!-- Zircaloy 4 -->
<material id="3">
<density value="6.55" units="g/cc" />
<nuclide name="O-16" xs="71c" wo="0.00125" />
<nuclide name="Nb-93" xs="71c" wo="0.01" />
<nuclide name="Sn-112" xs="71c" wo="0.000097" />
<nuclide name="Sn-114" xs="71c" wo="0.000066" />
<nuclide name="Sn-115" xs="71c" wo="0.000034" />
<nuclide name="Sn-116" xs="71c" wo="0.001454" />
<nuclide name="Sn-117" xs="71c" wo="0.000768" />
<nuclide name="Sn-118" xs="71c" wo="0.002422 " />
<nuclide name="Sn-119" xs="71c" wo="0.000859" />
<nuclide name="Sn-120" xs="71c" wo="0.003258" />
<nuclide name="Sn-122" xs="71c" wo="0.000463" />
<nuclide name="Sn-124" xs="71c" wo="0.000579" />
<nuclide name="Fe-54" xs="71c" wo="0.00005845" />
<nuclide name="Fe-56" xs="71c" wo="0.00091754" />
<nuclide name="Fe-57" xs="71c" wo="0.00002119" />
<nuclide name="Fe-58" xs="71c" wo="0.00000282" />
<nuclide name="Zr-90" xs="71c" wo="0.503052375" />
<nuclide name="Zr-91" xs="71c" wo="0.10970355" />
<nuclide name="Zr-92" xs="71c" wo="0.167684125" />
<nuclide name="Zr-94" xs="71c" wo="0.16993295" />
<nuclide name="Zr-96" xs="71c" wo="0.027377" />
</material>
<!-- Water -->
<material id="4">
<density value="0.73" units="g/cc" />
<nuclide name="H-1" xs="71c" ao="2.0" />
<nuclide name="O-16" xs="71c" ao="1.0" />
<sab name="lwtr" xs="04t" />
</material>
<!-- Guide Tube -->
<material id="5">
<density value="7.92" units="g/cc" />
<nuclide name="Ni-58" xs="71c" wo="0.06467315" />
<nuclide name="Ni-60" xs="71c" wo="0.02492135" />
<nuclide name="Ni-61" xs="71c" wo="0.001082905" />
<nuclide name="Ni-62" xs="71c" wo="0.00345287" />
<nuclide name="Ni-64" xs="71c" wo="0.000879225" />
<nuclide name="Fe-54" xs="71c" wo="0.04062275" />
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<nuclide name="Fe-57" xs="71c" wo="0.01472705" />
<nuclide name="Fe-58" xs="71c" wo="0.0019599" />
<nuclide name="Mn-55" xs="71c" wo="0.020" />
<nuclide name="Cr-50" xs="71c" wo="0.0082555" />
<nuclide name="Cr-52" xs="71c" wo="0.1591991" />
<nuclide name="Cr-53" xs="71c" wo="0.0180519" />
<nuclide name="Cr-54" xs="71c" wo="0.0044935" />
</material>
</materials>

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<?xml version="1.0"?>
<plots>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<plot id="1" color="mat" >
<origin>0. 0. 0.</origin>
<width>25.50 25.50 </width>
<pixels>1000 1000 </pixels>
<col_spec id="1" rgb="123 123 231" />
</plot>
</plots>

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<?xml version="1.0"?>
<settings>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<cross_sections>
/home/ahmedk/Desktop/xdata/cross_sections.xml
</cross_sections>
<eigenvalue>
<batches>600</batches>
<inactive>50</inactive>
<particles>10000</particles>
</eigenvalue>
<source>
<space>
<type>box</type>
<parameters>
-10.659 -10.659 -182.9
10.659 10.659 182.9
</parameters>
</space>
</source>
</settings>

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@ -0,0 +1,47 @@
<?xml version="1.0"?>
<tallies>
<tally id="1">
<filter type="cell" bins="1" />
<scores>flux</scores>
</tally>
<tally id="2">
<filter type="cell" bins="1" />
<scores>nu-fission</scores>
</tally>
<tally id="3">
<filter type="cell" bins="1" />
<scores>delayed-nu-fission</scores>
</tally>
<tally id="4">
<filter type="cell" bins="1" />
<scores>fission</scores>
</tally>
<tally id="5">
<filter type="cell" bins="1" />
<scores>absorption</scores>
</tally>
<tally id="6">
<filter type="cell" bins="1" />
<scores>scatter</scores>
</tally>
<tally id="7">
<filter type="cell" bins="1" />
<scores>total</scores>
</tally>
<tally id="8">
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
<scores>delayed-nu-fission</scores>
</tally>
</tallies>

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AP1000/PWR 2.6/summary.h5 Normal file

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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 46.5121 +/- 1.02981E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 1.29545 +/- 6.77359E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 9.15835E-03 +/- 4.46167E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.526995 +/- 2.77877E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00188 +/- 4.34804E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 31.8304 +/- 8.61290E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 32.8323 +/- 8.96590E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.65238E-04 +/- 1.40824E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.45796E-03 +/- 7.36344E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.43462E-03 +/- 7.15562E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 4.19965E-03 +/- 2.04970E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.35014E-03 +/- 6.15735E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.50745E-04 +/- 2.12997E-07

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<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
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<materials>
<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
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<!--
====================================================
Description: AP1000 PWR Assembly 17*17
Case: Problem 2 (17*17 PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
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BIN
AP1000/PWR 3.10/summary.h5 Normal file

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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 45.2682 +/- 9.83436E-03
============================> TALLY 2 <============================
Cell 1
Total Material
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Total Material
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Total Material
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============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00157 +/- 4.39105E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 30.2793 +/- 7.79164E-03
============================> TALLY 7 <============================
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Total Reaction Rate 31.2809 +/- 8.14277E-03
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Total Material
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Delayed Group 2
Total Material
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Total Material
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Total Material
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Delayed Group 5
Total Material
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Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.63176E-04 +/- 2.21685E-07

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<?xml version="1.0"?>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<?xml version="1.0"?>
<materials>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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@ -0,0 +1,20 @@
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<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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@ -0,0 +1,35 @@
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<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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BIN
ARC700/CANDU 2.1/summary.h5 Normal file

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@ -0,0 +1,84 @@
============================> TALLY 1 <============================
Cell 93
Total Material
Flux 163.424 +/- 5.28121E-02
Cell 87
Total Material
Flux 0.0 +/- 0.0
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Total Material
Nu-Fission Rate 1.26106 +/- 6.40821E-04
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Total Material
Nu-Fission Rate 0.0 +/- 0.0
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Total Material
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Total Material
Delayed-Nu-Fission Rate 0.0 +/- 0.0
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Total Material
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Total Material
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Total Material
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Total Material
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Total Material
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Total Material
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Total Material
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Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.40181E-03 +/- 6.95039E-07
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Total Material
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Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 3.98570E-03 +/- 1.92860E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.24331E-03 +/- 5.70304E-07
Delayed Group 6
Total Material
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File diff suppressed because one or more lines are too long

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@ -0,0 +1,217 @@
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====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<materials>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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View file

@ -0,0 +1,20 @@
<?xml version="1.0"?>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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@ -0,0 +1,35 @@
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<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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@ -0,0 +1,47 @@
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BIN
ARC700/CANDU 2.6/summary.h5 Normal file

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@ -0,0 +1,84 @@
============================> TALLY 1 <============================
Cell 93
Total Material
Flux 157.157 +/- 4.94838E-02
Cell 87
Total Material
Flux 0.0 +/- 0.0
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Cell 93
Total Material
Nu-Fission Rate 1.32360 +/- 6.54682E-04
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Total Material
Nu-Fission Rate 0.0 +/- 0.0
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Cell 93
Total Material
Delayed-Nu-Fission Rate 9.08830E-03 +/- 4.28404E-06
Cell 87
Total Material
Delayed-Nu-Fission Rate 0.0 +/- 0.0
============================> TALLY 4 <============================
Cell 93
Total Material
Fission Rate 0.540215 +/- 2.68485E-04
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Total Material
Fission Rate 0.0 +/- 0.0
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Total Material
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Total Material
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Total Material
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Total Material
Scattering Rate 0.0 +/- 0.0
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Total Material
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Total Material
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Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.72592E-04 +/- 1.35988E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.46917E-03 +/- 7.09705E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.43818E-03 +/- 6.88963E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 4.17091E-03 +/- 1.96860E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.29643E-03 +/- 5.81586E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.41012E-04 +/- 2.03381E-07

File diff suppressed because one or more lines are too long

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@ -0,0 +1,217 @@
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<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<?xml version="1.0"?>
<materials>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<default_xs>71c</default_xs>
<material id="11">
<density value="9.82" units="g/cc" />
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<nuclide name="U-238" xs="73c" wo="96.9" />
<nuclide name="O-16" xs="73c" wo="13.4529" />
</material>
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<nuclide name="Dy-163" xs="72c" wo="1.861" />
<nuclide name="Dy-164" xs="72c" wo="2.1317" />
</material>
<!-- Zircaloy 4 -->
<material id="2">
<density value="7.48" units="g/cc" />
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<nuclide name="Cr-50" xs="71c" wo="0.004345" />
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</materials>

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@ -0,0 +1,20 @@
<?xml version="1.0"?>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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</plot>
</plots>

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<?xml version="1.0"?>
<settings>
<!--
====================================================
Description: ACR1000 Assembly 43 Fuel pins
Case: Problem (43 fuel pins)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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/home/ahmedk/Desktop/xdata/cross_sections1.xml
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</parameters>
</space>
</source>
</settings>

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<?xml version="1.0"?>
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<scores>delayed-nu-fission</scores>
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<tally id="4">
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</tally>
</tallies>

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ARC700/CANDU 3.1/summary.h5 Normal file

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============================> TALLY 1 <============================
Cell 93
Total Material
Flux 152.257 +/- 4.58838E-02
Cell 87
Total Material
Flux 0.0 +/- 0.0
============================> TALLY 2 <============================
Cell 93
Total Material
Nu-Fission Rate 1.36990 +/- 7.02126E-04
Cell 87
Total Material
Nu-Fission Rate 0.0 +/- 0.0
============================> TALLY 3 <============================
Cell 93
Total Material
Delayed-Nu-Fission Rate 9.38572E-03 +/- 4.57395E-06
Cell 87
Total Material
Delayed-Nu-Fission Rate 0.0 +/- 0.0
============================> TALLY 4 <============================
Cell 93
Total Material
Fission Rate 0.559204 +/- 2.88087E-04
Cell 87
Total Material
Fission Rate 0.0 +/- 0.0
============================> TALLY 5 <============================
Cell 93
Total Material
Absorption Rate 1.00287 +/- 4.35912E-04
Cell 87
Total Material
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============================> TALLY 6 <============================
Cell 93
Total Material
Scattering Rate 54.6965 +/- 1.73526E-02
Cell 87
Total Material
Scattering Rate 0.0 +/- 0.0
============================> TALLY 7 <============================
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Total Material
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Cell 87
Total Material
Total Reaction Rate 0.0 +/- 0.0
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.82242E-04 +/- 1.45970E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.51899E-03 +/- 7.59665E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.48638E-03 +/- 7.36856E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 4.30767E-03 +/- 2.10212E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.33549E-03 +/- 6.16825E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 4.54956E-04 +/- 2.16568E-07

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<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
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<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<materials>
<default_xs>71c</default_xs>
<!-- Fuel composed of U-238 O-16 and U-235 2.1% -->
<material id="1">
<density value="10.55" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.0185105302348" />
<nuclide name="U-238" xs="80c" wo="0.862943290471" />
<nuclide name="O-16" xs="80c" wo="0.118546179294" />
</material>
<!-- 3.6% UO2 + 5% Gd2O3 -->
<material id="5">
<density value="10.547" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.0292792" />
<nuclide name="U-238" xs="80c" wo="0.7940422" />
<nuclide name="O-16" xs="80c" wo="0.1194922" />
<nuclide name="Gd-157" xs="80c" wo="0.0571954 " />
</material>
<!-- Cladding material alloy Э110 -->
<material id="2">
<density value="6.4516" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Hf-174" xs="80c" wo="0.00000048" />
<nuclide name="Hf-176" xs="80c" wo="0.00001578" />
<nuclide name="Hf-177" xs="80c" wo="0.0000558" />
<nuclide name="Hf-178" xs="80c" wo="0.00008184" />
<nuclide name="Hf-179" xs="80c" wo="0.00004086" />
<nuclide name="Hf-180" xs="80c" wo="0.00010524" />
<nuclide name="Zr-90" xs="80c" wo="0.50920065" />
<nuclide name="Zr-91" xs="80c" wo="0.11104434" />
<nuclide name="Zr-92" xs="80c" wo="0.16973355" />
<nuclide name="Zr-94" xs="80c" wo="0.17200986" />
<nuclide name="Zr-96" xs="80c" wo="0.0277116" />
</material>
<!-- Water -->
<material id="3">
<density value="0.716" units="g/cc" />
<nuclide name="H-1" xs="80c" wo="0.06312" />
<nuclide name="O-16" xs="80c" wo="0.13539" />
<nuclide name="B-11" xs="80c" wo="0.80149" />
</material>
<!-- Guide Tube alloy Э635 -->
<material id="4">
<density value="6.55" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Sn-112" xs="80c" wo="0.0001261" />
<nuclide name="Sn-114" xs="80c" wo="0.0000858" />
<nuclide name="Sn-115" xs="80c" wo="0.0000442" />
<nuclide name="Sn-116" xs="80c" wo="0.0018902" />
<nuclide name="Sn-117" xs="80c" wo="0.0009984" />
<nuclide name="Sn-118" xs="80c" wo="0.0031486" />
<nuclide name="Sn-119" xs="80c" wo="0.0011167" />
<nuclide name="Sn-120" xs="80c" wo="0.0042354" />
<nuclide name="Sn-122" xs="80c" wo="0.0006019" />
<nuclide name="Sn-124" xs="80c" wo="0.0007527" />
<nuclide name="Fe-54" xs="80c" wo="0.00017535" />
<nuclide name="Fe-56" xs="80c" wo="0.00275262" />
<nuclide name="Fe-57" xs="80c" wo="0.00006357" />
<nuclide name="Fe-58" xs="80c" wo="0.00000846" />
<nuclide name="Zr-90" xs="80c" wo="0.50662815" />
<nuclide name="Zr-91" xs="80c" wo="0.11048334" />
<nuclide name="Zr-92" xs="80c" wo="0.16887605" />
<nuclide name="Zr-94" xs="80c" wo="0.17114086" />
<nuclide name="Zr-96" xs="80c" wo="0.0275716" />
</material>
</materials>

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<?xml version="1.0"?>
<plots>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<plot basis="xy" id="1" color="mat" >
<origin>0. 0. 0.</origin>
<width>30.50 30.50 </width>
<pixels>1000 1000 </pixels>
<col_spec id="5" rgb="123 123 231" />
</plot>
<plot basis="yz" id="2" color="mat" >
<origin>0. 0. 0.</origin>
<width>50.50 50.50 </width>
<pixels>1000 1000 </pixels>
</plot>
</plots>

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<?xml version='1.0' encoding='utf-8'?>
<settings>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<cross_sections>
/home/ahmedk/Desktop/xdata/cross_sections1.xml
</cross_sections>
<eigenvalue>
<batches>600</batches>
<inactive>50</inactive>
<particles>10000</particles>
</eigenvalue>
<source>
<space>
<type>point</type>
<parameters>
0.0 0.0 0.0
</parameters>
</space>
</source>
</settings>

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<?xml version="1.0"?>
<tallies>
<tally id="1">
<filter type="cell" bins="1" />
<scores>flux</scores>
</tally>
<tally id="2">
<filter type="cell" bins="1" />
<scores>nu-fission</scores>
</tally>
<tally id="3">
<filter type="cell" bins="1" />
<scores>delayed-nu-fission</scores>
</tally>
<tally id="4">
<filter type="cell" bins="1" />
<scores>fission</scores>
</tally>
<tally id="5">
<filter type="cell" bins="1" />
<scores>absorption</scores>
</tally>
<tally id="6">
<filter type="cell" bins="1" />
<scores>scatter</scores>
</tally>
<tally id="7">
<filter type="cell" bins="1" />
<scores>total</scores>
</tally>
<tally id="8">
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
<scores>delayed-nu-fission</scores>
</tally>
</tallies>

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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 64.9018 +/- 1.33778E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 0.997313 +/- 5.50758E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 7.33525E-03 +/- 3.68047E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.404201 +/- 2.25497E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00166 +/- 4.10541E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 28.7426 +/- 5.85780E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 29.7443 +/- 6.10826E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.25891E-04 +/- 1.25118E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.22589E-03 +/- 6.50769E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.20457E-03 +/- 6.25374E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 2.83474E-03 +/- 1.42311E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.30311E-03 +/- 6.16517E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 5.41041E-04 +/- 2.56891E-07

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<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<geometry>
<cell fill="5" id="1" region="1 -2 -80 110 -70 100 -120 90" />
<cell id="101" material="void" region="-5" universe="909" />
<cell id="102" material="1" region="5 -6" universe="909" />
<cell id="500" material="void" region="6 -7" universe="909"/>
<cell id="600" material="2" region="7 -8" universe="909" />
<cell id="100" material="3" region="8" universe="909" />
<cell id="111" material="void" region="-5" universe="709" />
<cell id="112" material="5" region="5 -6" universe="709" />
<cell id="510" material="void" region="6 -7" universe="709"/>
<cell id="610" material="2" region="7 -8" universe="709" />
<cell id="110" material="3" region="8" universe="709" />
<cell id="107" material="3" region="-9" universe="808" />
<cell id="108" material="4" region="9 -10" universe="808" />
<cell id="109" material="3" region="10" universe="808" />
<cell id="103" material="3" region="-11" universe="505" />
<cell id="105" material="4" region="11 -12" universe="505" />
<cell id="106" material="3" region="12" universe="505" />
<cell id="601" material="3" universe="900" />
<hex_lattice id="5" n_rings="11">
<pitch>1.275 </pitch>
<outer>900</outer>
<center>0.0 0.0 </center>
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909
</universes>
</hex_lattice>
<surface id="1" type="z-plane" coeffs="-176.5" boundary="reflective" />
<surface id="2" type="z-plane" coeffs="176.5" boundary="reflective" />
<surface id="70" type="plane" coeffs="1 1.73205 0 23.6" boundary="reflective"/>
<surface id="80" type="x-plane" coeffs="11.8" boundary="reflective" />
<surface id="90" type="plane" coeffs="-1 1.73205 0 -23.6" boundary="reflective" />
<surface id="100" type="plane" coeffs="1 1.73205 0 -23.6 " boundary="reflective" />
<surface id="110" type="x-plane" coeffs="-11.8 " boundary="reflective" />
<surface id="120" type="plane" coeffs="-1 1.73205 0 23.6" boundary="reflective" />
<surface id="5" type="z-cylinder" coeffs="0. 0. 0.075" />
<surface id="6" type="z-cylinder" coeffs="0. 0. 0.3785" />
<surface id="7" type="z-cylinder" coeffs="0. 0. 0.3865" />
<surface id="8" type="z-cylinder" coeffs="0. 0. 0.455" />
<surface id="9" type="z-cylinder" coeffs="0. 0. 0.545" />
<surface id="10" type="z-cylinder" coeffs="0. 0. 0.628" />
<surface id="11" type="z-cylinder" coeffs="0. 0. 0.55" />
<surface id="12" type="z-cylinder" coeffs="0. 0. 0.63" />
</geometry>

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<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<materials>
<default_xs>71c</default_xs>
<!-- Fuel composed of U-238 O-16 and U-235 2.6% -->
<material id="1">
<density value="10.55" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.0229176280555" />
<nuclide name="U-238" xs="80c" wo="0.858529604849" />
<nuclide name="O-16" xs="80c" wo="0.118552767096" />
</material>
<!-- 3.6% UO2 + 5% Gd2O3 -->
<material id="5">
<density value="10.547" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.0292792" />
<nuclide name="U-238" xs="80c" wo="0.7940422" />
<nuclide name="O-16" xs="80c" wo="0.1194922" />
<nuclide name="Gd-157" xs="80c" wo="0.0571954 " />
</material>
<!-- Cladding material alloy Э110 -->
<material id="2">
<density value="6.4516" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Hf-174" xs="80c" wo="0.00000048" />
<nuclide name="Hf-176" xs="80c" wo="0.00001578" />
<nuclide name="Hf-177" xs="80c" wo="0.0000558" />
<nuclide name="Hf-178" xs="80c" wo="0.00008184" />
<nuclide name="Hf-179" xs="80c" wo="0.00004086" />
<nuclide name="Hf-180" xs="80c" wo="0.00010524" />
<nuclide name="Zr-90" xs="80c" wo="0.50920065" />
<nuclide name="Zr-91" xs="80c" wo="0.11104434" />
<nuclide name="Zr-92" xs="80c" wo="0.16973355" />
<nuclide name="Zr-94" xs="80c" wo="0.17200986" />
<nuclide name="Zr-96" xs="80c" wo="0.0277116" />
</material>
<!-- Water -->
<material id="3">
<density value="0.716" units="g/cc" />
<nuclide name="H-1" xs="80c" wo="0.06312" />
<nuclide name="O-16" xs="80c" wo="0.13539" />
<nuclide name="B-11" xs="80c" wo="0.80149" />
</material>
<!-- Guide Tube alloy Э635 -->
<material id="4">
<density value="6.55" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Sn-112" xs="80c" wo="0.0001261" />
<nuclide name="Sn-114" xs="80c" wo="0.0000858" />
<nuclide name="Sn-115" xs="80c" wo="0.0000442" />
<nuclide name="Sn-116" xs="80c" wo="0.0018902" />
<nuclide name="Sn-117" xs="80c" wo="0.0009984" />
<nuclide name="Sn-118" xs="80c" wo="0.0031486" />
<nuclide name="Sn-119" xs="80c" wo="0.0011167" />
<nuclide name="Sn-120" xs="80c" wo="0.0042354" />
<nuclide name="Sn-122" xs="80c" wo="0.0006019" />
<nuclide name="Sn-124" xs="80c" wo="0.0007527" />
<nuclide name="Fe-54" xs="80c" wo="0.00017535" />
<nuclide name="Fe-56" xs="80c" wo="0.00275262" />
<nuclide name="Fe-57" xs="80c" wo="0.00006357" />
<nuclide name="Fe-58" xs="80c" wo="0.00000846" />
<nuclide name="Zr-90" xs="80c" wo="0.50662815" />
<nuclide name="Zr-91" xs="80c" wo="0.11048334" />
<nuclide name="Zr-92" xs="80c" wo="0.16887605" />
<nuclide name="Zr-94" xs="80c" wo="0.17114086" />
<nuclide name="Zr-96" xs="80c" wo="0.0275716" />
</material>
</materials>

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<?xml version="1.0"?>
<plots>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<plot basis="xy" id="1" color="mat" >
<origin>0. 0. 0.</origin>
<width>30.50 30.50 </width>
<pixels>1000 1000 </pixels>
<col_spec id="5" rgb="123 123 231" />
</plot>
<plot basis="yz" id="2" color="mat" >
<origin>0. 0. 0.</origin>
<width>50.50 50.50 </width>
<pixels>1000 1000 </pixels>
</plot>
</plots>

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<?xml version='1.0' encoding='utf-8'?>
<settings>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<cross_sections>
/home/ahmedk/Desktop/xdata/cross_sections1.xml
</cross_sections>
<eigenvalue>
<batches>600</batches>
<inactive>50</inactive>
<particles>10000</particles>
</eigenvalue>
<source>
<space>
<type>point</type>
<parameters>
0.0 0.0 0.0
</parameters>
</space>
</source>
</settings>

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<?xml version="1.0"?>
<tallies>
<tally id="1">
<filter type="cell" bins="1" />
<scores>flux</scores>
</tally>
<tally id="2">
<filter type="cell" bins="1" />
<scores>nu-fission</scores>
</tally>
<tally id="3">
<filter type="cell" bins="1" />
<scores>delayed-nu-fission</scores>
</tally>
<tally id="4">
<filter type="cell" bins="1" />
<scores>fission</scores>
</tally>
<tally id="5">
<filter type="cell" bins="1" />
<scores>absorption</scores>
</tally>
<tally id="6">
<filter type="cell" bins="1" />
<scores>scatter</scores>
</tally>
<tally id="7">
<filter type="cell" bins="1" />
<scores>total</scores>
</tally>
<tally id="8">
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
<scores>delayed-nu-fission</scores>
</tally>
</tallies>

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============================> TALLY 1 <============================
Cell 1
Total Material
Flux 63.7640 +/- 1.32958E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 1.05453 +/- 5.91572E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 7.70342E-03 +/- 3.90170E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.427644 +/- 2.42306E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00180 +/- 4.29561E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 28.0157 +/- 5.83095E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 29.0175 +/- 6.09326E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.38933E-04 +/- 1.34371E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.29291E-03 +/- 6.96071E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.26839E-03 +/- 6.66953E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 2.97715E-03 +/- 1.50882E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.36080E-03 +/- 6.41808E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 5.65231E-04 +/- 2.67859E-07

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<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<geometry>
<cell fill="5" id="1" region="1 -2 -80 110 -70 100 -120 90" />
<cell id="101" material="void" region="-5" universe="909" />
<cell id="102" material="1" region="5 -6" universe="909" />
<cell id="500" material="void" region="6 -7" universe="909"/>
<cell id="600" material="2" region="7 -8" universe="909" />
<cell id="100" material="3" region="8" universe="909" />
<cell id="111" material="void" region="-5" universe="709" />
<cell id="112" material="5" region="5 -6" universe="709" />
<cell id="510" material="void" region="6 -7" universe="709"/>
<cell id="610" material="2" region="7 -8" universe="709" />
<cell id="110" material="3" region="8" universe="709" />
<cell id="107" material="3" region="-9" universe="808" />
<cell id="108" material="4" region="9 -10" universe="808" />
<cell id="109" material="3" region="10" universe="808" />
<cell id="103" material="3" region="-11" universe="505" />
<cell id="105" material="4" region="11 -12" universe="505" />
<cell id="106" material="3" region="12" universe="505" />
<cell id="601" material="3" universe="900" />
<hex_lattice id="5" n_rings="11">
<pitch>1.275 </pitch>
<outer>900</outer>
<center>0.0 0.0 </center>
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909 909 909 909 808 909 909 909 909 909 909
909 909 909 909 909 909 909 909 909 909
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</universes>
</hex_lattice>
<surface id="1" type="z-plane" coeffs="-176.5" boundary="reflective" />
<surface id="2" type="z-plane" coeffs="176.5" boundary="reflective" />
<surface id="70" type="plane" coeffs="1 1.73205 0 23.6" boundary="reflective"/>
<surface id="80" type="x-plane" coeffs="11.8" boundary="reflective" />
<surface id="90" type="plane" coeffs="-1 1.73205 0 -23.6" boundary="reflective" />
<surface id="100" type="plane" coeffs="1 1.73205 0 -23.6 " boundary="reflective" />
<surface id="110" type="x-plane" coeffs="-11.8 " boundary="reflective" />
<surface id="120" type="plane" coeffs="-1 1.73205 0 23.6" boundary="reflective" />
<surface id="5" type="z-cylinder" coeffs="0. 0. 0.075" />
<surface id="6" type="z-cylinder" coeffs="0. 0. 0.3785" />
<surface id="7" type="z-cylinder" coeffs="0. 0. 0.3865" />
<surface id="8" type="z-cylinder" coeffs="0. 0. 0.455" />
<surface id="9" type="z-cylinder" coeffs="0. 0. 0.545" />
<surface id="10" type="z-cylinder" coeffs="0. 0. 0.628" />
<surface id="11" type="z-cylinder" coeffs="0. 0. 0.55" />
<surface id="12" type="z-cylinder" coeffs="0. 0. 0.63" />
</geometry>

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@ -0,0 +1,84 @@
<?xml version='1.0' encoding='utf-8'?>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<materials>
<default_xs>71c</default_xs>
<!-- Fuel composed of U-238 O-16 and U-235 3.7% -->
<material id="1">
<density value="10.55" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.03221694" />
<nuclide name="U-238" xs="80c" wo="0.84921578" />
<nuclide name="O-16" xs="80c" wo="0.11856726" />
</material>
<!-- 3.6% UO2 + 5% Gd2O3 -->
<material id="5">
<density value="10.547" units="g/cc" />
<nuclide name="U-235" xs="80c" wo="0.0292792" />
<nuclide name="U-238" xs="80c" wo="0.7940422" />
<nuclide name="O-16" xs="80c" wo="0.1194922" />
<nuclide name="Gd-157" xs="80c" wo="0.0571954 " />
</material>
<!-- Cladding material alloy Э110 -->
<material id="2">
<density value="6.4516" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Hf-174" xs="80c" wo="0.00000048" />
<nuclide name="Hf-176" xs="80c" wo="0.00001578" />
<nuclide name="Hf-177" xs="80c" wo="0.0000558" />
<nuclide name="Hf-178" xs="80c" wo="0.00008184" />
<nuclide name="Hf-179" xs="80c" wo="0.00004086" />
<nuclide name="Hf-180" xs="80c" wo="0.00010524" />
<nuclide name="Zr-90" xs="80c" wo="0.50920065" />
<nuclide name="Zr-91" xs="80c" wo="0.11104434" />
<nuclide name="Zr-92" xs="80c" wo="0.16973355" />
<nuclide name="Zr-94" xs="80c" wo="0.17200986" />
<nuclide name="Zr-96" xs="80c" wo="0.0277116" />
</material>
<!-- Water -->
<material id="3">
<density value="0.716" units="g/cc" />
<nuclide name="H-1" xs="80c" wo="0.06312" />
<nuclide name="O-16" xs="80c" wo="0.13539" />
<nuclide name="B-11" xs="80c" wo="0.80149" />
</material>
<!-- Guide Tube alloy Э635 -->
<material id="4">
<density value="6.55" units="g/cc" />
<nuclide name="Nb-93" xs="80c" wo="0.0100" />
<nuclide name="Sn-112" xs="80c" wo="0.0001261" />
<nuclide name="Sn-114" xs="80c" wo="0.0000858" />
<nuclide name="Sn-115" xs="80c" wo="0.0000442" />
<nuclide name="Sn-116" xs="80c" wo="0.0018902" />
<nuclide name="Sn-117" xs="80c" wo="0.0009984" />
<nuclide name="Sn-118" xs="80c" wo="0.0031486" />
<nuclide name="Sn-119" xs="80c" wo="0.0011167" />
<nuclide name="Sn-120" xs="80c" wo="0.0042354" />
<nuclide name="Sn-122" xs="80c" wo="0.0006019" />
<nuclide name="Sn-124" xs="80c" wo="0.0007527" />
<nuclide name="Fe-54" xs="80c" wo="0.00017535" />
<nuclide name="Fe-56" xs="80c" wo="0.00275262" />
<nuclide name="Fe-57" xs="80c" wo="0.00006357" />
<nuclide name="Fe-58" xs="80c" wo="0.00000846" />
<nuclide name="Zr-90" xs="80c" wo="0.50662815" />
<nuclide name="Zr-91" xs="80c" wo="0.11048334" />
<nuclide name="Zr-92" xs="80c" wo="0.16887605" />
<nuclide name="Zr-94" xs="80c" wo="0.17114086" />
<nuclide name="Zr-96" xs="80c" wo="0.0275716" />
</material>
</materials>

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@ -0,0 +1,28 @@
<?xml version="1.0"?>
<plots>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<plot basis="xy" id="1" color="mat" >
<origin>0. 0. 0.</origin>
<width>30.50 30.50 </width>
<pixels>1000 1000 </pixels>
<col_spec id="5" rgb="123 123 231" />
</plot>
<plot basis="yz" id="2" color="mat" >
<origin>0. 0. 0.</origin>
<width>50.50 50.50 </width>
<pixels>1000 1000 </pixels>
</plot>
</plots>

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@ -0,0 +1,33 @@
<?xml version='1.0' encoding='utf-8'?>
<settings>
<!--
====================================================
Description: VVER 1000 (Russian PWR Design)
Case: Problem 3 (Hexagonal PWR Assembly)
Written By: Ahmed K Madani
Date: 02/03/2017
====================================================
-->
<cross_sections>
/home/ahmedk/Desktop/xdata/cross_sections1.xml
</cross_sections>
<eigenvalue>
<batches>600</batches>
<inactive>50</inactive>
<particles>10000</particles>
</eigenvalue>
<source>
<space>
<type>point</type>
<parameters>
0.0 0.0 0.0
</parameters>
</space>
</source>
</settings>

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@ -0,0 +1,46 @@
<?xml version="1.0"?>
<tallies>
<tally id="1">
<filter type="cell" bins="1" />
<scores>flux</scores>
</tally>
<tally id="2">
<filter type="cell" bins="1" />
<scores>nu-fission</scores>
</tally>
<tally id="3">
<filter type="cell" bins="1" />
<scores>delayed-nu-fission</scores>
</tally>
<tally id="4">
<filter type="cell" bins="1" />
<scores>fission</scores>
</tally>
<tally id="5">
<filter type="cell" bins="1" />
<scores>absorption</scores>
</tally>
<tally id="6">
<filter type="cell" bins="1" />
<scores>scatter</scores>
</tally>
<tally id="7">
<filter type="cell" bins="1" />
<scores>total</scores>
</tally>
<tally id="8">
<filter type="delayedgroup" bins="1 2 3 4 5 6" />
<scores>delayed-nu-fission</scores>
</tally>
</tallies>

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BIN
VVER/VVER 3.6/summary.h5 Normal file

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63
VVER/VVER 3.6/tallies.out Normal file
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@ -0,0 +1,63 @@
============================> TALLY 1 <============================
Cell 1
Total Material
Flux 61.9896 +/- 1.28230E-02
============================> TALLY 2 <============================
Cell 1
Total Material
Nu-Fission Rate 1.13900 +/- 6.24803E-04
============================> TALLY 3 <============================
Cell 1
Total Material
Delayed-Nu-Fission Rate 8.24555E-03 +/- 4.13592E-06
============================> TALLY 4 <============================
Cell 1
Total Material
Fission Rate 0.462219 +/- 2.55841E-04
============================> TALLY 5 <============================
Cell 1
Total Material
Absorption Rate 1.00181 +/- 4.24571E-04
============================> TALLY 6 <============================
Cell 1
Total Material
Scattering Rate 26.8902 +/- 5.57261E-03
============================> TALLY 7 <============================
Cell 1
Total Material
Total Reaction Rate 27.8920 +/- 5.83435E-03
============================> TALLY 8 <============================
Delayed Group 1
Total Material
Delayed-Nu-Fission Rate 2.58189E-04 +/- 1.41929E-07
Delayed Group 2
Total Material
Delayed-Nu-Fission Rate 1.39177E-03 +/- 7.36533E-07
Delayed Group 3
Total Material
Delayed-Nu-Fission Rate 1.36245E-03 +/- 7.06298E-07
Delayed Group 4
Total Material
Delayed-Nu-Fission Rate 3.18686E-03 +/- 1.59938E-06
Delayed Group 5
Total Material
Delayed-Nu-Fission Rate 1.44552E-03 +/- 6.80326E-07
Delayed Group 6
Total Material
Delayed-Nu-Fission Rate 6.00761E-04 +/- 2.83970E-07

81
chain_simple.xml Normal file
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@ -0,0 +1,81 @@
<?xml version="1.0"?>
<depletion_chain>
<nuclide name="I135" decay_modes="1" reactions="1" half_life="2.36520E+04">
<decay type="beta" target="Xe135" branching_ratio="1.0" />
<reaction type="(n,gamma)" Q="0.0" target="Xe136" /> <!-- Not precisely true, but whatever -->
</nuclide>
<nuclide name="Xe135" decay_modes="1" reactions="1" half_life="3.29040E+04">
<decay type=" beta" target="Cs135" branching_ratio="1.0" />
<reaction type="(n,gamma)" Q="0.0" target="Xe136" />
</nuclide>
<nuclide name="Xe136" decay_modes="0" reactions="0" />
<nuclide name="Cs135" decay_modes="0" reactions="0" />
<nuclide name="Gd157" decay_modes="0" reactions="1" >
<reaction type="(n,gamma)" Q="0.0" target="Nothing" />
</nuclide>
<nuclide name="Cs137" half_life="949252600.0" decay_modes="2" decay_energy="179447.6443" reactions="6">
<decay type="beta-" target="Ba137" branching_ratio="0.05300549"/>
<decay type="beta-" target="Ba137_m1" branching_ratio="0.94699451"/>
<source type="discrete" particle="photon">
<parameters>283500.0 4.235177914969822e-15</parameters>
</source>
<source type="discrete" particle="electron">
<parameters>513971.0 892130.0 1175630.0 6.915023250821416e-10 4.235177914969822e-15 3.870076370575872e-11</parameters>
</source>
<reaction type="(n,2n)" Q="-8278100.0" target="Cs136"/>
<reaction type="(n,3n)" Q="-15047200.0" target="Cs135"/>
<reaction type="(n,p)" Q="-3562520.0" target="Xe137"/>
<reaction type="(n,a)" Q="3064210.0" target="I134"/>
<reaction type="(n,gamma)" Q="4281390.0" target="Cs138" branching_ratio="0.9021"/>
<reaction type="(n,gamma)" Q="4281390.0" target="Cs138_m1" branching_ratio="0.0979"/>
</nuclide>
<nuclide name="Ba137" reactions="5">
<reaction type="(n,2n)" Q="-6903090.0" target="Ba136"/>
<reaction type="(n,3n)" Q="-16015200.0" target="Ba135"/>
<reaction type="(n,gamma)" Q="8611390.0" target="Ba138"/>
<reaction type="(n,p)" Q="-390530.0" target="Cs137"/>
<reaction type="(n,a)" Q="6046200.0" target="Xe134"/>
</nuclide>
<nuclide name="Ba137_m1" half_life="153.12" decay_modes="1" decay_energy="661396.64" reactions="0">
<decay type="IT" target="Ba137" branching_ratio="1.0"/>
<source type="discrete" particle="photon">
<parameters>4470.0 31817.0 32194.0 36304.0 36378.0 37255.0 661657.0 4.372683136122041e-05 9.525884763557368e-05 0.00017377297596020668 1.6633087848867745e-05 3.2112650589648466e-05 1.0158463048637348e-05 0.004069614128287558</parameters>
</source>
<source type="discrete" particle="electron">
<parameters>3670.0 26400.0 624216.4 655668.2 660364.2 661404.0 661636.9 0.0003537897992698539 3.7210487357645964e-05 0.00037210482830822686 6.708752100247119e-05 1.4350025269251869e-05 3.08729347663194e-06 4.6173597451093537e-07</parameters>
</source>
</nuclide>
<nuclide name="Gd156" decay_modes="0" reactions="1">
<reaction type="(n,gamma)" Q="0.0" target="Gd157" />
</nuclide>
<nuclide name="U234" decay_modes="0" reactions="1">
<reaction type="fission" Q="191840000."/>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>1.093250e-04 2.087260e-04 2.780820e-02 6.759540e-03 2.392300e-02 4.356330e-05</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
<nuclide name="U235" decay_modes="0" reactions="1">
<reaction type="fission" Q="193410000."/>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>6.142710e-5 1.483250e-04 0.0292737 0.002566345 0.0219242 4.9097e-6</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
<nuclide name="U238" decay_modes="0" reactions="1">
<reaction type="fission" Q="197790000."/>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Gd157 Gd156 I135 Xe135 Xe136 Cs135</products>
<data>4.141120e-04 7.605360e-04 0.0135457 0.00026864 0.0024432 3.7100E-07</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
</depletion_chain>

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