Boron Neutron Capture Therapy(BNCT) is one of the efficient method to treat some cancer tumors, particularly the brain tumors(glioblastoma multiform). In this treatment boron with particular chemical compound is concentrated within the tumor. Then a high intensity beam of epithermal neutrons is focused on the tumor area on the head. Thermal neutron in the reaction 10 B(n,?) 7 Li produces energetic alpha and lithium particles with a range about the size of the cells. These charged particles deposit their kinetic energies to the cell which leads to destroy the cancer cell. For boron neutron capture therapy we need a high flux of thermal neutrons at the place of tumor, because the cross section of the 10 B(n,?) 7 Li reaction has a high value for thermal neutrons. Before neutrons reach the tumor, they should go through the skin, skull and brain tissues. Most of thermal neutrons in the beam are captured by the normal cells. So they can produce damages. The epithermal neutrons slow down by elastic or inelastic collisions. Therefore thermal neutrons should be removed and fast neutrons should be moderated to convert epithermal neutrons. O the other hand the ?-rays from inelastic collisions should be shielded to protect their damages on the normal cells. Therefore neutron beam to use in BNCT should have some characteristics: ? epi 5×10 9 n/cm 2 s, D nf / ? epi 2×10 -13 Gycm 2 , D ? / ? epi 2×10 -13 Gycm 2 , ? epi/ ? th 20, j/? 0.7. For this purpose a Beam Shaping Assembly(BSA) should be designed. Designing a beam shaping assembly related to the kind of neutron source. To design a BSA many parameters should be determined: kind of materials, shape and dimentions of materials. Usually 6 Li is used for capturing the thermal neutrons, AlF 3 is used as moderator of the fast neutrons, Bi is used for filtering the ?-rays and Pb is used as reflector and collimator of neutrons. One of the high intensity neutron source in Isfahan is Miniature Neutron Source Reactor(MNSR). This reactor is a pool tank type reactor with power of 30kW, highly enriched uranium fuel, natural water moderator and metal beryllium reflector. Its neutro flux is i order of 10 10 - 10 12 n/cm 2 .s. We have simulated the neutron traort from the core of the reactor to the output windows of BSA, for different design of BSA. For this purpose we have used the MCNP4C code which is a suitable Monte Carlo code to simulate the neutron interactions in different materials and geometries. Our calculations show that the neutron flux of the MNSR reactor can be used for BNCT and we have presented a design of a BSA to produce optimum neutron beam with required characteristics for BNCT. Keywords: Neutron therapy, BNCT, MNSR, Monte Carlo method