Missing nuclide in low energy neutron library

Dear Fluka experts,

I’m new to FLUKA. From my ADS problem, I am creating geometry and fill in with desired materials but the “Low energy neutron xsec not found for some media” error has occurred with this list of materials
So I have followed this Low energy neutron xsec not found for some media forum and I have found that the list of materials that I mentioned aren’t listed in low energy neutron library.

I have thought of 2 solutions to figure out this error :
1.simulate my problem to high energy neutron (>20 MeV) like calculate the recoil as mentioned in this lecture https://www.fluka.org/free_download/course/portugal2010/Lectures/AdvancedMaterial2010.pdf
but I’m not sure how to do it and I don’t know it will give the expected result or not?

2.add another library by converting the ENDF file to group-wise cross-section to FLUKA (I don’t know is it available or not?)

Please correct my opinion

Thank you in advance,

Dear @pattarapol.13

Some low-energy neutron cross sections are implemented only for the natural isotopic composition of the material, i.e. FLUKA will give you an error if you choose a single isotope (see comment in this post). You might get lucky if you use simply use the predefined materials.

You can implement this workaround with the PART-THR card or you can discard neutrons all together with the DISCARD card if they are not relevant for your problem.

I believe that it is not possible to change or choose the low-energy neutron library in FLUKA (see also this post), but maybe someone from the developer team might give you some more details about that.


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Dear @lorenzo.mercolli,

Thank you for your explanations. I have some questions to make sure that I am understanding.
If I’m not miss understanding, the post said that don’t need to specify others isotope because it is already included in natural isotopic composition. In my case, I want to use Pu-239 which is listed in low energy neutrons library, but I still want other isotopes of Pu. I didn’t see any natural isotopic composition of Pu. So that mean I couldn’t use others isotope?

Best regard,

Hi Thanapong

There is no ‘natural’ composition of Pu since its entirely artificial, anything higher than Z>92 is not considered natural. I don’t think FLUKA handles other isotopes of Pu.



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Dear @thanapong

As you have understood, it is not presently possible to accurately model the bahaviour of your nuclear fuel mixture due to the absence of low-energy neutron cross sections for various (minor) actinide isotopes. If your goal is to do detailed studies of the neutron population in your ADS reactor (k_eff studies etc.), this can be a significant drawback, further compounded by the group-wise treatment of neutrons below 20 MeV.

It is up to you to decide, based on the above, whether you can obtain the desired results from your FLUKA simulation (I am just making some assumptions about it). If not, codes like MCNP are intentionally tailored and therefore more commonly used for reactor/criticality studies. Note that you can very easily export your FLUKA geometry to MCNP via the export option in Flair, removing the need for duplication of your geometry-building effort.

Dear @atsingan
Thank you for your help. I have tried to use all materials existing and match to my problem. I got an error when I define density in “mixele” material by 9502.5 g/cm^3. I calculated this value from the summation of all fuel weights divided by the volume of fuel region. it gives an error attached below but when I change the density to 2.5(assume) it works fine. Could you give me a suggestion on how to calculate the density of material for the compound card?
ads_final.inp (3.0 KB) ads_final.out (660 Bytes)


Dear @pattarapol.13

A density of ~9500 g/cm3 is completely unrealistic by 2-3 orders of magnitude. No material of such density exists at normal temperature and pressure ranges, and indeed FLUKA fails to digest it. (Admittedly, the error message provided in the ads_final001.out file, rather than the ads_final.out which contains very little information, is not particularly informative.) Perhaps you meant to type 9.5 g/cm3, which would not be unreasonable given the composition of the mixele compound.

Determining the density of a mixture based on the density of its components can be tricky. If the assumption holds that the volume of the mixture is equal to the sum of the volumes of the components, then, given the densities d_i and mass abundances a_i of the components, the density d_mix of the mixture would be:

d_mix = Σ{a_i * d_i} / Σ{a_i}

where it may be that Σ{a_i} = 1, if the abundances are normalised.

For your compound material, this would give something around 10.51 g/cm3. Note that this value is dominated by the density of the Pb-Bi eutectic alloy (10.5 g/cm3, 89.9% abundance), which decreases visibly with increasing reactor temperature.