Dear experts,
Iam using MCNP (version 6.1) to extract neutron self-shielding factors for slab-shaped geometries and a fast neutron spectrum as input (energy range from 60 keV to 20 MeV). In the simulations my samples are irradiated by a parallel beam of neutrons, with the same cross section as the sample front surface, so its completely irradiated.
The self-shielding factor should be obtained by dividing the result of the F4 tally in my sample with the material (density) inside by the F4 tally result with void (no material inside). However, I have realised that for small thicknesses (e.g. at 1 mm) the resulting self-shielding factor would be slightly above 1 (I think that should not be possible). For me it seems like this might come from down-scattering events, which are still in the considered tally energy bin range (60 keV to 20 MeV). In other publications I have never seen factors above 1 and they partly used even much smaller samples.
Does anyone have an idea or hint what I might do wrong in my simulations?
Thanks in advance.
Kind regards,
Niklas
Hi Niklas,
Correct me if I am wrong; isn't the self shielding the ratio of the mean fluence rate inside the sample volume to the fluence rate incident on the sample? With that being said, how the statistics are looking? especially when you are calculating the ratio. And what material are you analyzing in your MCNP calcs? I am asking to see if you expect to "lose" neutrons when they pass through that.
Regards,
dimitris
Dear Dimitrios,
thank you very much for your reply!
You are right, I derive the factor in MCNP as you have said. One has to divide the result of the F4 tally in the sample volume filled with the material and its nominal physical density by the same F4 tally result but infinitely diluted (that means the sample is now filled with void, ~ 0 g cm^-3 density). Your definition with the surface flux is equivalent; for a void sample of constant geometric cross section and a parallel beam of particles the surface flux is also constant for every thickness (particle track length equals the thickness of the sample in this case, divided by the volume yields the surface flux).
The integral statistics (60 keV to 20 MeV) are totally sufficient, something like 0.01% statistical error only.
Basically four our fast neutrons the total macroscopic cross sections equal the scattering cross section, i.e. they only scatter. In that case it is logical that 1 primary neutron can make more than 1 score in the F4 tally, if its energy after interaction still is in the range between 60 keV and 20 MeV and if it is still inside the sample. Therefore, meanwhile I would assume MCNP is correct and. By the way, the shielding factors are not much above 1, mostly something below 1.05.
Iam analyzing different materials, severals metals like aluminum and iron or some compounds like calcium carbonate (CaCO3) or cerium chloride (CeCl3).
Kind regards,
Niklas
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