Hi Thomas,
You could use mgdraw.f routine to print out individual particle crossings with the logic of your choice.
I can imagine that you could latch neutrons so that you don’t count the same twice through the same boundary.
-M
On Mar 12, 2019, at 9:29 AM, Thomas Smith <thomas.smith-14_at_student.manchester.ac.uk<mailto:thomas.smith-14_at_student.manchester.ac.uk>> wrote:
Dear Fluka experts,
I am trying to find the attenuation in the number of thermal neutrons when passing through different shielding materials and would like the know the number of neutrons at given intermediate thicknesses within the material. I was wondering whether there was a simple way to determine the population of neutrons which entered a given region or the population of neutrons crossing a boundary such that if a single neutron is scattered multiple times into the region or across the boundary, it is not counted more than once. So far, I have been using a series of input files, each containing a given thickness of the shielding material equal to an intermediate thickness and have been using USRYIELD to determine the number entering and exiting the region containing the material. It would be more convenient to do this using one input file containing many regions whbilst also not over-counting the number of neutrons in each region and was wondering whether this is possible.
Best regards,
Thomas
__________________________________________________________________________
You can manage unsubscription from this mailing list at
https://www.fluka.org/fluka.php?id=acc_info
Received on Wed Mar 13 2019 - 04:20:53 CET