I’m trying to calculate the neutron source (as a function of energy) due to (alpha,n) reactions in uranium fuel, similarly to the SOURCES4C code. I am simulating an alpha-source using the U235 and U238 decay spectrum of alpha to weight the source particles. The geometry is essentially a homogenous region of uranium fuel, surrounded by the black hole region.
I just want to check I’m setting up the file correctly. In order to output the neutron flux (in neutrons per second per cm^3), I need to use USRTRACK or USRCOLL? Does the detector volume have to match the volume of the region defined in USRTRACK/USRCOLL?
Also I notice the results in the sum.lis file are in Part/GeV/cmq/pr, which I’m guessing is particles per GeV per cm^3 but what is the ‘pr’?
I have attached my input files. In the sum.lis file I’m getting 0 results for all energy groups, so I’m not sure what I’m doing wrong.
USRTRACK and USRCOLL score similar quantities: both of them give an energy spectrum of the particle fluence. USRTRACK scores it through the track length (therefore the fluence itself), while USRCOLL exploits the reaction rate to calculate it. The limitation of the latter is the fact that it yields always 0 in vacuum. Look also here:
Pr stands for primary particle, notice that is per cm^2, not cube. All results in FLUKA are always given per primay particle in general.
You are simulating an uncommon reaction channel. I suggest to use biasing to enhance neutron production. To do so, we need to bias the inelastic length of the alpha particle in your material.
A quite strong biasing could be (just to observe the neutron spectra):
LAM-BIAS 0.0000001 MFuel 4-HELIUM
Please make sure to read also the biasing lectures present in the past courses, since I see that you have already a weight set in your source file. Also, considering the nature of your simulations, I strongly advise in using LOW-PWXS card when there is an interest in following low energy neutrons.
Let me know if that works for you!
Cheers,
Daniele
Thank you very much for your help. I’ve implemented the biasing and low energy neutron cross-sections for the fuel region but I’m still getting zero results for the USRTRACK and USRCOLL estimators. I tried increasing the size of the detector but this didn’t make a difference. snre_fuel_v4.flair (2.5 KB) snre_fuel_v4.inp (2.4 KB)
Indeed. I took a more careful look at your inputfile and I tried to run these simulations myself. The first thing I noticed is the fact that you are not declaring a DEFAULTS card. Then, FLUKA automatically considers NEW-DEFAults. While sometimes these are acceptable, they set: Particle transport threshold set at 10 MeV, except for neutrons (1E-5 eV)
This would not happily coexist with your simulation, since your primary particles already start below 10 MeV. You can either change this threshold manually via PART-THRes card or (better) use a more precise defaults card, as PRECISIOn.
Secondly, your material composition and your alpha particle energy are quite an unfortunate combination. You have chosen to include in your simulation Zr(nat), C-12, U-235, U-238. Your maximum energy in the file is 4.598 MeV. I retrieved the cross section of (HE4, n) reaction, and none of the material can produce neutrons in the exit channel:
This is due to the specificity of your isotopic composition of the materials. If, for instance, you were considering natural carbon instead of C-12, the C-13 component has a higher cross-section, and you would observe some neutrons.
Hope that this solves the mystery!
Cheers,
Daniele
I have included the DEFAULTS card with PRECISIOn and also corrected the graphite to natural carbon so it should include C13. I am still getting zero results, apart from the lower energy groups which are close to zero. Am I correct in thinking that I don’t need the LOW-NEUT card if I have activated the DEFAULTS card?
With the point-wise treatment of neutrons, we do not have the concept of low energy groups. I tried to run your simulation, and I see that most of your neutrons are indeed the MeV range. They do not thermalize due to the specificity of your simulation.
Here there are my results (obtained with your inputfile and a logarithmic scoring). Do you obtain similar ones?
You can also avoid the usage of the LOW-NEUT card since it is automatically turned on by many defaults. Also the LOW-MAT card is not necessary when using the pointwise treatment of neutrons.
which shows that all neutrons have energies less than 10MeV. I haven’t quite worked out plotting yet as I’m still fairly new to FLUKA. Correct me if I’m wrong but are you saying that the results are invalid due to the specific problem I am trying to model?
Sorry, I’m not expecting neutrons above 10MeV. I was confused by the statement ‘With the point-wise treatment of neutrons, we do not have the concept of low energy groups’. Does this mean that using the point-wise treatment of neutrons is not valid for low energy groups?
On the contrary, the point-wise treatment is henceforth the recommended option for low-energy neutrons. @dcalzola was simply pointing out that in this case referring to energy groups* is (by definition) an inappropriate wording, without questioning the result validity. By the way, the latter here does not depend mainly on the low-energy neutron transport, rather on their production through the alpha reaction model.
*not to be confused with the energy binning applying in general to spectrum scoring
