In this thesis, the Isfahan Miniature Neutron Source Reactor (MNSR) was simulated using the MCNP code, and the relative neutron flux distribution and also ?-ray dose distribution along the length of the dry channel of this reactor, which is placed next to the core of this reactor were calculated. The above quantities were also determined experimentally by the method of Neutron Activation Analysis (NAA) and by application of ?-ray pocket dosimeters respectively. The calculated results were then compared with the corresponding measured values and showed good agreements, which indicate that the designed simulation program is accurate enough to be used for various kinds of calculations throughout of this reactor including determination of the neutron flux and ?-ray dose at locations where experimental measurements are not possible. In addition, the designed simulation program was used to determine the neutron energy spectrum in the dry channel and in the inner and outer irradiation sites of this reactor. The thermal parts of these spectrums were then compared with the expected Maxwellian distribution.