Study of a neutron radiography system using 252Cf neutron source

This paper presents a study about moderation and collimation of a neutron radiography system using 252Cf. A Monte Carlo Code, MCNP4B, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plane. Among the various moderator materials investigated, high density polyethylene proved to be the most efficient, with a thermalization factor of 56 cm2. Using a collimator design assembly it was possible to obtain a normalized thermal neutron flux, at the image plane of 6x10-6 n cm-2 s-1 at an effective collimator ratio of 7.5, or 3.2x10-7 n cm-2 s-1 at an effective collimator ratio of 50