Hello Fluka admins and users,
This is my first post for 2021, so happy new year!
I am performing some neutron yield benchmarking calculations with Fluka. I am trying to compare the neutron yields of different primary & target configurations. I am considering protons and deuterons of different energies (in the MeV range). I am testing the neutron yield when either of these beams are incident on beryllium targets and lithium targets of appropriate thicknesses (as obtained from SRIM).
I have a rather simple geometry under consideration as shown in my input file. I am looking at the neutron flux in different directions as denoted by my USRBDX cards. I am scoring neutrons of all energies, thermal neutrons and epithermal neutrons in each angular range. Please note that my definition of thermal and epithermal neutrons might be smaller than you might define it because it is a neutron scientist’s definition of those energy ranges.
I have a few questions and I am listing them from hard/complicated to easy/trivial:
1.) When I utilize deuterons of small energies (2 MeV), I get no neutrons scored in all angular ranges. I am unsure why this is the case as deuterons should undergo a reaction with both beryllium and lithium targets. When I gradually increase this energy above 6MeV, I get some neutrons scored, but the errors are huge (~25%). I believe there is something inherently wrong with the model that I am using which is causing my simulation to catastrophically fail. Is there a physics card that I am missing or need to deactivate to correct my model?
2.) I obtain results for protons on lithium and beryllium with reasonable errors. However, the results are only as reliable as my model. I want to ensure that my model is appropriate. If any feedback can be provided on the cards which I use to simulate me problem, I will appreciate that.
3.a) I want to understand, in greater detail, under what circumstances LOW-MAT should be activated.
The manual simply states that beginner users (like myself) can misinterpret this. However, as low energy neutron transport is extremely important for my applications, I need to have a firm grasp of its utility. In my case, I activated LOW-MAT for the target material.
3 b.) However is it also important to assign a cross-section to the medium, water, which appears in my geometry and is being used to moderate/slow the neutrons coming from the target? If I am indeed supposed to assign a LOW-MAT property to water, how do I go about doing this?
3 c.) What specific properties does Fluka modify when the LOW-MAT card is called for a given element? I have a collaborator who uses MCNP for example, and he has to use different data tables depending on the target material (Li vs Be) and energy regime of primary particles considered. I basically want to ensure that the quantities used by Fluka are appropriate for my simulations. Also, if I wanted to check the numerical values of the data (e.g. cross-section) which Fluka uses, how can I do this?
4.) As you can see, my USRBDX card scores the two-way flux at each boundary crossing. I have also included the area of the boundary crossing in each USRBDX card. When I look at my sum.lis file, I get the total response in each detector (units of particles/cm^2/primary and particles/primary). I can convert this to a flux by considering the beam current and proton (or deuteron) charge. Is there any additional normalization factors which I need to include in order to obtain the two-way neutron flux crossing this boundary?
FYI, my simulations are run with 10,000,000 primaries, 10 MC cycles, and each simulation is also spawned 10 times to improve the statistics.
I’ve provided two input files corresponding to different simulations and three sum.lis files per simulation - each of the sum.lis files contains scoring of thermal, epithermal and neutrons of all energies across each of the detectors defined for different neutron scattering angles (relative to the z-axis/propagation axis).
Thanks for any insight you can provide.
DaliniBe_deut_2MeV_0pt5.inp (11.3 KB) Be_deut_2MeV_0pt5_21_sum.lis (1.4 MB) Be_deut_2MeV_0pt5_22_sum.lis (1.3 MB) Be_deut_2MeV_0pt5_23_sum.lis (1.5 MB) Be_deut_8MeV_0pt5.inp (11.3 KB) Be_deut_8MeV_0pt5_22_sum.lis (1.3 MB) Be_deut_8MeV_0pt5_22_tab.lis (605.2 KB) Be_deut_8MeV_0pt5_23_sum.lis (1.5 MB)