MCNP program is used to calculate the neutron transmission coefficient of 3–9 cm-thick neutron absorber material B4C/Al composite with 5%–15% B4C content in air, water, 200–1400 ppm (1 ppm=10-6) H3BO3 solution, irradiated by 0.5–20 MeV neutrons and 235U thermal neutron fission source. The results show that the transmission coefficient of B4C/Al composite decreases with the increase of the content of B4C and the thickness of material, but increases with the increase of neutron energy, and has little influence from the variation in H3BO3 solution concentration. A better shielding effect of B4C/Al composite is displayed in water than in H3BO3 solution, and a “reversal” phenomenon of the shielding effect occurs in air. The neutron transmission coefficient is almost unchanged with neutron energy when neutron energy is higher than the 5–15 MeV. The neutron transmission coefficient of B4C/Al composite irradiated under a fission source is lower than under a steady 20 MeV neutron source. Ranking the shielding performances of media, the sequence is H3BO3 solution > water > air, and the exponential decay relationships between neutron transmission coefficient and thickness of medium can be expressed as e-0.71x and e-0.669x, where x is thickness of medium in cm.