Dosimetric Quantities and Spectra of Neutrons Transmitted Through and Reflected from Different Homogeneous Slabs

摘要

Spectra of neutrons transmitted through and reflected from different homogeneous slabs are calculated by the Monte Carlo code 05R5S and by the albedo code MUSPALB. Their dosimetric quantities, i.e. average reaction cross section of some activation detectors and average values per unit fluence for tissue kerma, absorbed dose, dose equivalent and gamma-ray dose are calculated by the code SPECTRANS. The spectra tabulated in this paper were published in the form of plots in an earlier report. (Atomindex citation 09:356542)