| Neutron activation analysis is one of the main methods to analyze elemental composition and content in the cement, alloys, etc. It has developed into a fully mature technology, and has been widely used in the social production and people's daily lives. For example, industrial raw material components usually need to be analyzed, especially in the case of real-time monitoring, which is difficult to finish by other methods. For Fe, Ca, Si, S, etc, the thermal neutron capture reaction cross section both are very large, so it is easy to get the characteristic peak. Therefore, a high-intensity thermal neutron source is needed in the neutron activation analysis of mineral composition. The energy of neutrons emitted by common neutron source is in the fast neutron energy range, so it needs to be slowed into thermal neutrons. 252Cf neutron source is an isotope neutron source, which is double-shell package, the inner layer with 90% Platinum and 10% rhodium; the outer layer is packaged with stainless steel or zirconium alloy. The average energy of neutrons emitted by 252Cf neutron source is 2.348MeV, and its activity reaches 2.35×1012n/s per gram. So it is the most promising radioisotope neutron source. In this paper, the 252Cf fission neutron source with high activity is selected. In the practical applications, a reasonable moderator material and structure need to be selected to slow the fast neutrons into the thermal neutrons for improving the thermal neutron fluency rate. Monte Carlo method is widely used to solve the problem of neutron transport today, the structure of moderator and shield of 252Cf neutron source is studied with MCNP 4C program.Monte Carlo Method can vividly describe the random events and simulate the experiment procedure, and it is very suitable to solve particle transport problem which has the randomness. The method has been used to simulate nuclear experiments widely, and it can simulate various micro-physical processes by using the known reaction cross section data. MCNP 4C is made by Los Alamos Station Lab in USA based on the Monte Carlo thoughts, and it can deal with neutron-photon coupled transport of three-dimensional geometric structure.In this paper the following two aspects are simulated with Monte Carlo method. Firstly, polyethylene, water, heavy water, graphite, lead are simulated to get the capability of slowing fast neutron. The structure of moderator is established initially according to the characteristics of different materials. Secondly, the structure of shielding is analyzed by MCNP 4C program. The neutron disclosed and the gamma-ray emitted are considered for the protection issue of 252Cf neutron source. The size of shielding materials is calculated according to the radiation protection standards, and the appropriate thickness of the shielding is choosed so that the total dose surrounding the system does not exceed the value of protection standards.To improve the fluency rate of thermal neutron and meet the standard of radiation protection, the structure of moderator and shield of 252Cf neutron source is designed as follows. Polyethylene is used as the first layer to slow the fast neutrons into the thermal neutrons; Graphite reflector is selected as the second layer to improve fluency rate of thermal neutron; Paraffin containing 10% boron is used as the third layer to absorb the neutron disclosed from the system; iron is used as the outermost layer to reduce dose equivalent rate of gamma rays to the limit of radiation protection standard.
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