ZENG Xinmiao ZHOU Peng QIN Peizhong BAO Mao GUO Guangshui XU Ziyan. Monte Carlo simulation of neutron transmission through various materials[J]. Nuclear techniques, 2011, 34(3): 188-192.
ZENG Xinmiao ZHOU Peng QIN Peizhong BAO Mao GUO Guangshui XU Ziyan. Monte Carlo simulation of neutron transmission through various materials[J]. Nuclear techniques, 2011, 34(3): 188-192.DOI:
Monte Carlo simulation of neutron transmission through various materials
the MCNP code is used to calculate neutron transmission through various materials of 1–40 cm thickness in energy range of 0.025 eV to 20 MeV.The materials are principal shielding materials of PE
water
concrete
Fe
Cu
Pb and PE/BC4composite
and the additives of B4C and Gd2O3.The following results were obtained:(1) the neutron transmission decreased with increasing thicknesses of a material;(2) for a certain material
with decreasing energy of the neutrons
the neutron transmission decreased faster
except PE and water
where the neutron transmision keeps almost the same in the entire energy range;and(3) for low energy neutrons
the differences of transmission decreasing trend with the increasing thicknesses for different materials were significant
whereas for high energy neutron
the decreasing trend was not distinct.A combination of a main material with an additives can achieve better shielding effect.