Dear Fluka colleagues
Have a good day
In the field of neutron interaction with matter, today I wanted to talk about what I encountered in the interaction of thermal neutrons definitely at energy 25.4 meV with materials. The goal was of collecting and counting neutrons at a gate outside the reaction side. I found the following: taking the material as a whole and comparing to take it as isotopes. I found that the calculations differ greatly. For this reason I spoke with colleagues who use MCNP. They informed me that MCNP the new version MCNP6 uses this property to distribute down matter into its isotopic elements. So, as an example, I used as example; gadolinium as consisting of 6 isotopes as in the following figure Gd-154 - Gd-160
I found that the result is completely identical to what was published before . I would like to open discussion with who uses the material as bulk or divided into its isotopes fractions especially in case of neutron interaction with matter. And why it is different from bulk or the isotope fraction modes. I ask permission for open discussion with colleague how face and have experience in such calculations!
Best Regards
Mohamed Fayez
Hi @mohammed.fayiz
When you request natural element in FLUKA, it will use the isotopic composition predefined in the code, and will ask the neutron database (selected with the LOW-PWXS) for the isotope. If the isotopes are not existing it will try to load the natural one if it exists. The new databases like JEFF-3.3, JENDL-4.0, ENDF-VIII only contain isotopes and not natural elements (with the exception of Carbon).
You can verify in the output the isotope composition that was requested for Gadolinium, in the section for neutrons
*** Low energy neutron Point Wise materials
### Material Z A m T(K)/Abu Dataset Ver Filename
26 GADOLINI 64 0 0 296.0K JEFF-3.3 1.1 64_0_Gadolinium
158 0 0.2484
160 0 0.2186
156 0 0.2047
157 0 0.1565
155 0 0.1480
154 0 0.0218
152 0 0.0020
Which matches exactly with what you have requested (except the Gd-152)
However one strange thing I see in your screenshot is that you have asked the S(a,b) cross section to be loaded for Aluminum metal?
There is no S(a,b) data for Gadolinium in our database, binding to Aluminium will be wrong.
Many thanks receiving this kind answer.
1- As informed it is option to request the natural element or the form of isotopic structures. My problem was I couldn’t gain the same results for either way. Well, they are very close to each other but not identical. Moreover requesting isotopic compositions values gives a very correct and near result compared with the published one. Do you think we should recommend this for all neutron interaction works?
2- Another topic: To obtain accurate results for some calculations, the target thickness must be taken close and in the order of (1/10) of the mean free path MFP (cm). So please I ask is the value of MFP included in FLUKA output file or not? Or how to request its value for any requested reaction?
3- Thank you very much again. Correcting me the mistake of including “Al” without any meaning.
You have the scattering lengths printed in the output per material for the beam particle, but not for the low-energy neutrons (E<20MeV).
There are many web sites with endf browsers where you can see the cross section per isotope, from which you can calculate the MFP.
