This study investigates shielding design of an isotropic 241Am-9Be neutron source using Monte Carlo Code MCNPX. Typical Am-Be neutron sources emit neutrons with a broad spectrum. Different materials were studied in terms of both moderating power of first layer and absorbing ability of the second one. This arrangement is consistent with safety requirements, cost limitations and material availability. After optimizing the moderator thickness by MCNP code, different materials for attenuating neutrons, most of which were thermal because of moderating, were examined. Then moderator was fixed and the best shield configuration was chosen to minimize equivalent dose outside the shield. For this purpose, MCNPX flux to dose conversion factor was used. Finally, proper sites were determined in order to achieve maximum thermal, epithermal and fast neutron flux. This configuration enables us to use neutron flux of sites with different energy ranges for irradiating samples without exposing personnel to radiation.