I am currrently trying to benchmark results for neutron yields obtained in FLUKA and MCNP. In my simulation, a spherical beryllium target (radius 0.7mm) is centred at the origin and is bombarded by a 10MeV proton source. This interaction between the protons and the target produces neutrons.
I defined a spherical detector (5 cm radius) which is also centered on the origin (as shown in the above figure). I computed the one-way neutron current across this detector surface and I integrated over all solid-angle by invoking the scoring quantity I1,logE,lin(omega) and assigning one angular bin.
I would like to know if there is anyone with experience in MCNP and knows whether the F2 tally in MCNP is an equivalent quantity to compare against the results obtained from scoring neutrons in the way described above in FLUKA.
Thanks for your reply! This has saved me a lot of headache.
Can you also please further clarify whether the one-way (I1) or two-way (I2) current in FLUKA is the equivalent to the F1 tally?
Also, thanks for also making the additional note re: equivalence of fluence and current if I score a point-like quantity. However, I do not, and in some simulations I use moderator materials, in which the neutrons bounce around in the moderating medium. Therefore, in the moderator case, these quantities are quite different.
f1 scores two-way current, but you can split contributions from both directions with the cosine card, so that it will score one-way in each cosine bin: *c1 90 0
However, I do not think you need this with your geometry since you don’t have particles entering your sphere from outside.