A study of a neutron radiography system using 252Cf neutron source

This paper presents a study about moderation, collimation and shielding of a transportable neutron radiography system utilizing californium-252. A Monte Carlo Code, MCNP3b, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plane, considering the geometric unsharpness, and an adequate radiation shielding to attend radiological protection rules. Among the various moderator materials investigated, the high density polyethylene showed to be the most efficient, with a thermalization factor of 56 cm2 . With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plane, an L/D ratio 7.5, for neutron flux up to 6.10-6 cm-2 .s-1 per neutron source. The total dose equivalent rates were significantly reduced by the shielding optimization process. (author)