Energy filters of neutrons in Flux calculation

User Support
1

Hello, I am doing flux calculation in openMC, I want to calculate for three energy groups or bins.

0 - 0.625 eV
0.625 eV–100 keV
100 keV–20 MeV

How to code this in xml? Right now, my tallies.xml is

<?xml version="1.0"?>

regular

120 1

<lower_left> -60 -60 </lower_left>

1 120

1

1

flux

Thanks for the help.

2

Hello,

You want to add an energy filter to your tallies.xml file and then update your filters array to include this new filter.

The energy filter would be:

  <filter id="2" type="energy">
    <bins>0. 0.625 100.0e3 20.0e6</bins>
  </filter>

The updated filters array within the <tally> would be:
  <filters>1 2</filters>
where the 2 refers to the filter id of the energy filter you want.

For an example like this, see the examples/xml/pincell/ folder

3

Hello,

What is the default energy if I did not specify the energy filter?

4

If no energy filter is used, then a particle of any energy will contribute to the tally, not only those within the range of energies listed in the filter. That is, no filtering on energy is done at all.