Dear Junyu
There is no straightforward way for users to add cross sections used for neutron transport (multigroup) below 20 MeV. Note that if you want to compare what is used in FLUKA with other libraries, you have the possibilities to print the cross section information in the output file by changing the default value of WHAT(4) in the LOW-NEUT card (WHAT(4) = 4.0 will give you the maximum including residual nuclei production cross sections). Don’t do it for your production runs as it will duplicate the information in different cycles and create large output files…
Also if you use RESNUCLE for calculation the production yield, you can change the value of the WHAT(1) to distinguish the contribution from neutrons below 20 MeV (cross sections) from the rest (models).
Finally, another approach is to calculate fluence spectra that you can fold with production cross section of interest and weight by the atomic density of the target nuclei. Note that in that case the shape of the spectra is still determined with cross sections built-in in the code, so this approach is maybe more relevant for production on trace elements having a more limited influence on the shape of the spectra.
Hoping this help.
Joachim
From: owner-fluka-discuss_at_mi.infn.it [mailto:owner-fluka-discuss_at_mi.infn.it] On Behalf Of Junyu Zhang
Sent: 10 January 2018 02:35
To: fluka-discuss_at_fluka.org
Subject: [fluka-discuss]: How can I apply my own cross section data in FLUKA
Dear FLUKA experts:
In my simulation I need to find out the yield of positron emitter like C-11, O-15 etc. but the default cross section data in FLUKA seems not exactly right, it is a little different from the cross section data in ENDF reported in ICRU-63. How can I apply this data to take place of the FLUKA default data?
Junyu
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Received on Wed Jan 10 2018 - 11:09:44 CET