Estimation of Self-Shielding Effect for RI Production

The thermal neutron flux distributions inside and outside cylindrical target materials of different sizes and different neutron absorption cross sections have been determined to provide information necessary for deciding the shape and arrangement of target materials for radioisotope production. A geometrical function independent of the size and cross section of the target material is presented, which approximates the neutron flux distribution experimentally obtained. Calculations have been made for hollow cylinders of infinitely thin wall and for thin disks. For such geometry the neutron flux decrease is proportional to the solid angle projected between source and target. The experimental values have been found practically in agreement with the values calculated by the formulas. The self-shielding effect of most kinds of targets for radioisotope production can well be estimated approximately by using the formulas presented.