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  • Shikhar Kumar
  • openmc
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  • #1220

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Opened Apr 11, 2019 by Shikhar Kumar@shikhark
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Wrong criticality when first nuclide in material has density of 0

Created by: ohnemax

This is a very specific error, and will very likely not affect many people. And probably it is an example for a wrongly written input file. However, it affected me and it took me some time to figure out what was going on, so I post it here.

When the first nuclide in a material has a density set to 0, criticality results show a very erratic behavior. A working example is attached.

Basically, the following material composition yields a combined k-effective of 0.76571 +/- 0.00025:

<material id="1" name="1.6% Fuel">
    <temperature>900</temperature>
    <density units="g/cm3" value="10.31341" />
    <nuclide ao="0.0" name="U234" />
    <nuclide ao="0.00037503" name="U235" />
    <nuclide ao="0.022625" name="U238" />
    <nuclide ao="0.046007" name="O16" />
  </material>

However, if the line with ao="0.0" for U234 is removed, it works fine. In this case criticality is 1.01150 +/- 0.00024. If the line is moved somewhere after a nuclide that has a non-zero density, it works fine, too. (criticality 1.01150).

I tested this with v0.10.0, compiled for cluster usage using OpenMPI 2.0.2. It occurs both for ENDF-B VII.0 and ENDF-B VIII cross section data sets.

Example Files Example with wrong result Example with correct result

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Reference: shikhark/openmc#1220