Dear Fluka experts,
After the flair gives .out file, it shows the n/p in the summary tab. However, average energy seems not shown in this section. It needed to be calculated by the user? or is there somewhere else to find these values (average energy and Fraction of high-energy neutrons)?
Dear @thanapong,
Thanks for your question.
As you said, you can obtain the neutron yield (neutron per primary) in the .out in the section:
âNumber of secondaries generated in inelastic interaction per particleâ.
Of course, It is only in the case that all your interactions are in one region (in your case, the material for the spallation)
In the cases of average energy of neutrons generated and fraction of high-energy neutrons, there is no way to obtain them from the .out file.
My suggestion to you is to use an extra scoring to obtain the yield of neutrons, and with this results you could calculate them.
You could obtain the three values using USRYIELD, using the EMERGING option, where you obtain âyields of particle emerging from inelastic hadronic interactions with single nucleiâ.
An example of how you should use this card is shown here:
Here you see that we use the quantities of kinetic energy and particle charge.
We defined explicitely that we want to score NEUTRONS in the input, but USRYIELD scores double-differential yield, then we had to use also the particle charge as a quantity, giving the range or -0.5 to 0.5.
You should just fill the range of energies of your problem.
Finally you will obtain an spectrum where you can calculate by yourself the neutron yield, average energy and fraction of high-energy neutrons.
Let me know if itâs clear, otherwise we can keep discussing.
Best.
André
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Thank you for your kind information. May I ask a little bit more about getting the spectrum to calculate the neutron yield, average energy, and fraction of high-energy neutrons. Is it in the .out file or do I need to obtain the result from .lis file by using usysuw ?.
Dear @thanapong,
What you say is correct.
But in case your are using Flair for your simulations,
in the process of merging your results, what you are actually doing is
using the routine USYSUW.
This routine merges the results and gives you 3 outputs (MANUAL, Note 13):
tab.lis, sum.lis)
It returns differential and cumulative fluence,
with the corresponding percent errors, in a file, and differential
fluence in another file formatted for easy plotting.
.yie)
It also returns a binary file that can be read out in turn by USYSUW.
As a quick check of your results, you can compare the results of neutron yield in
sum.lis â âTot. response (integrated over x1)â
The value obtained should be similar to the one you see in the .out files.
To conclude⊠yes, you have to use USYSUW to obtain the tab.lis and sum.lis
from where you can calculate by yourself the three results that you are looking for,
but these files are obtained automatically in the process of merging the results.
I hope this can help you,
Cheers,
André
@adonadon Thank you so much, I will try on it.
One more question about .out file. If I want the number of low energy neutron yield, I have to see at the column ânumber of secondaries created by low energy neutron per beamâ right? and What is the meaning of neutron and fission neutron in that column?
Thanks in advance.
Dear @thanapon,
Sorry for the delay in replying.
- If I want the number of low energy neutron yield, I have to see at the column ânumber of secondaries created by low energy neutron per beamâ right?
Thatâs correct, BUT what you are getting there is the number of low energy neutrons generated from the reinteractions of low energy neutron. Is this what you are looking for?
In order to define properly what you need, itâs important to define the following:
Are you interested in neutron production by proton spallation reactions or do you want to include also neutron reinteractions?
Answering this question will allow use to properly define what we want to score and how to do it.
Let me know.
- What is the meaning of neutron and fission neutron in that column?
Fission neutron in that column means âneutrons generated in fissions induced by low energy neutronsâ.
A bit info in the manual
Cheers,
Have a nice weekend
André
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Dear @adonadon ,
Thank you for your explanation.
Iâm interested in neutron production by proton spallation reactions.
Have a nice day,
Thanapong
Dear @thanapong,
Perfect. Then, my recommendation is to use the USRYIELD card discussed above, and in order to get only the neutron production by proton spallation reactions, you should define a thin target ( <= Inelastic scattering length).
Doing this you will get what you are looking for.
Extra comment: In case you donât know the scattering length just doing a simple run with 1 particle you can obtain it from the .out file.
Let me know if it has sense for you.
Have a nice day,
André