I'm using MCNP ssw and ssr card for neutron penetration analysis in reactor shielding material. Several source surfaces is generated in sequences starting from the nearest surface from the core up to the furthest surface (closer and closer to the tallies). I tallied the flux, both from KCODE only input, and with ssr card input. Does the discrepancy (even though it is relatively small) caused by the source surface distance from the core? Or the gap step of surfaces in each sequence calculation?
Try surrounding the tally with SSW surfaces in order to record as many neutron tracks as possible. You probably added NONU/TOTNU card? Usually I use imp:n=0 inside the volume around which SSW is recorded, so the backscattered neutrons are not recorded twice.
Hello, I am a new user on MCNP, I use the ssw map to obtain the phase space at the output of a proton accelerator. Initially generating 10000 particles, my wssa file only saves 125 particles. It seems incoherent to me. could someone help me?
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