source fixed

User Support
1

hi
Is there a way I can use OpenMC to do a neutron sheilding problem? I want to have an isotropic point source of neutrons, a paraffin shield, please can you give me please example of source fixed in the file .xml?

thank you

2

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3

Afternoon,

There is a way - shielding problems are typically solved with a fixed source calculation. Here, instead of iterating to solve the critical flux and neutron multiplication factor (k_eff), the problem simply takes a given source and transports it throughout the problem domain.

OpenMC can do these calculations; see the description of the <fixed_source> element in the documentation here. Use this instead of the element.

4

I should have added, an example settings.xml file for an isotropic point source of neutrons with energy 2MeV at an x,y,z location of <1.0,1.5,2.0> centimeters is shown below:

`


<?xml version="1.0"?>
<settings>

  <!-- Parameters for fixed source calculation -->
  <fixed_source>
    <batches>100</batches>
    <particles>10000</particles>
  </fixed_source>

  <!-- Source Definition -->
  <source>
    <space type="point">
      <parameters>1.0 1.5 2.0</parameters>
    </space>
    <!-- Note that this is optional as the default angular distribution is isotropic -->
    <angle>
      <type>isotropic</type>
    </angle>
    <energy>
      <type>monoenergetic</type>
      <energy>2.0</energy>
    </energy>

  </source>

</settings>

`

For more information on the source element go here

5

HI Hamyd,

  1. The fixed source is configured in settings.xml, as Adam Nelson mail also explains;

  2. The files are attached to the group mail, for the sake of completeness, and

  3. About ERROR.These input files will not work in your computer, because the cross-section I am using are not the same you are using and not installed the same way. (take a look at 3rd line of settings.xml)

Have fun,

Matheus

image.png

slab-561-dresner.zip (2.12 KB)