| B4C/Al composites use aluminum as the matrix material and B4C as the neutron absorbing material.Due to its good thermal neutron absorption performance,low density,good toughness,relatively high temperature resistance,radiation resistance,simple preparation process,and low raw material costs,B4C/Al composites have been widely studied by scholars worldwide,and been used as neutron shielding in the storage and transfer of reactor spent fuel.At present,the manufacturing technology of B4C/Al composites mainly includes metal melting casting,infiltration and powder metallurgy.Powder metallurgy is the most mature technology,which is also been widely used among the three manufacturing methods of B4C/Al composites.The neutron shielding performance of B4C/Al composites is one of the important parameters in practice.The main factors affecting the shielding performance of the B4C/Al composites are the content of B4C,10B abundance,density,B4C particle size and distribution,etc.This work has been introduced the process of manufacturing B4C/Al composites by powder metallurgy,and the measurement results of the material density,phase and micro morphology,which provides the possibility for the analysis and measurement of the neutron shielding performance of the material.The focus of this work is to analyze the factors affecting the neutron shielding performance of B4C/Al composites via means of theoretical analysis,Monte Carlo simulation and measurements,to provide reference and basis for the optimization and improvement of the material.For the application of B4C/Al composites in the neutron shielding of the fuel salt tank of molten salt reactor,it analyzed the neutron source,radioactivity intensity and neutron energy distribution of the fuel salt tank of molten salt reactor after the reactor been shutdown.By compared with cadmium and ordinary concrete materials,it analyzed the advantages of B4C/Al composites used as the neutron absorbing material of the molten salt reactor fuel salt tank.The neutron shielding proposal of the molten salt reactor fuel salt tank composed of B4C/Al composites and polyethylene has been determined through calculation.The contents of chapters in this thesis are as follows:Chapter one,introduction:As a new type of neutron absorbing material,B4C/Al composites have been widely studied by scholars,such as its preparation method,mechanical properties and radiation resistance.Factors affecting the neutron shielding performance of B4C/Al composites haven’t been systematically investigated and experimentally verified.This chapter summarizes the research status of B4C/Al composites,the limitation of current researchs,neutron shielding theory and analysis methods,the fesibility of B4C/Al composites compared to B4C and boron aluminum alloy materials,and the research carried out in this thesis for B4C/Al composites.Chapter two,sample manufacturing and materials properties of B4C/Al composites:The B4C/Al composites used in the experiment of this work were manufactured by powder metallurgy.The results of the density measurement of B4C/Al composites are close to the theoretical values,which indicates that the micropores inside the composite can be avoided and the density of the materials can be increased by cold pressing,hot pressing sintering,rolling process and so on.The X-ray diffractometer（XRD）has been used to measure the physical phase of B4C/Al composites,and no presence of boron and aluminum compounds in the material was found.It indicates that the probability of chemical reaction between aluminum and boron carbide is low.Scanning Electron Microscope（SEM）has been used to observe the microstructure morphology of the surface of B4C/Al composites sample,which provides a basis for the subsequent analysis of the difference between the numerical simulation and the experimental measurement results of the material’s neutron shielding performance.Through calculating the displacement per atom（DPA）of B4C/Al composites by neutron irradiation,the damage mechanism of B4C/Al composites been irradiated by neutron has been analyzed.It provides theoretical guidance for B4C/Al composites as a non-structural material for neutron shielding of fuel salt tank of molten salt reactor.Chapter three,the simulation analysis of neutron shielding performances of B4C/Al composites:The good neutron shielding performance is one of the important factors that B4C/Al composites are widely used as the neutron shielding material.Factors affecting the neutron shielding performance of B4C/Al composites are B4C content,10B abundance,material density and neutron flux.In this chapter,MCNP code,based on Monte Carlo method,has been used to calculate the shielding properties of B4C/Al composites for different energy neutrons according to the wide energy range of 10B（n,alpha）7Li reaction.It provides theoretical reference for shielding neutron adaptation range of B4C/Al composites and combined use with other neutron slowing materials.The thermal neutrons shielding performances of B4C/Al composites with various B4C contents,10B abundances and material densities has been analyzed to provide theoretical guidance for material preparation optimization.The influence to shielding performances of B4C/Al composites has been analyzed,which provides theoretical data for B4C/Al as a non-structural material that can be used in a high flux neutron irradiation environment for a long time.By comparing the calculation results of spherical structure and plate structure,it was confirmed that the contribution of neutron scattering leads to the slight difference between the results of simulation and theoretical calculation.Chapter four,the experimental measurements of neutron shielding performance of B4C/Al composites:The shielding performance of B4C/Al composites plates with16.85%and 31%boron carbide mass fraction in the energy range of 2×10-9Me V～5×10-4Me V was measured with the photo-neutron source driven by electron accelerator.The influence of particle size of B4C to the neutron shielding properties of B4C/Al composites has been analyzed by dividing B4C/Al composites into B4C particles and aluminum matrix.The results show that the increase of particle size of B4C powder affect the neutron shielding performances of B4C/Al composites,especially the low energy neutrons with energy less than 10-7Me V.It was found that if the particle size of B4C decreases to several microns,the neutron shielding performance of the composite would approach the theoretical level.This analysis result provides guidance for the selection of the particle size of the B4C raw material during the material manufacturing process.Chapter five,the application of B4C/Al composites in neutron shielding of fuel salt tank in molten salt reactor:The molten salt reactor uses Li F-Be F2-Zr F4-UF4（65.30 mol%-28.71 mol%-4.79 mol%-1.20 mol%）as fuel salt.After the reactor shutdown,gamma rays released by radionuclides in the fuel salt produce a large number of neutrons by（γ,n）reactions with nuclides such as beryllium,which is higher than the neutrons produced by the spontaneous fission of the fuel salt.The neutron source intensity and energy distribution through radionuclide decay,radionuclide fission and（α,n）reaction were calculated with ORIGEN-S code,and the neutron source intensity and energy distribution of the（γ,n）reaction between gamma ray and nuclides were calculated with MCNPX code.By analyzing the source intensity and energy distribution of the fuel salt neutrons after the reactor been shut down,the influence of neutron radiation around the fuel salt tank has been calculated.The B4C/Al composites have a neutron absorption ability in a wide energy range,which is significantly better than cadmium and concrete materials to shield the neutrons in the fuel salt tank.Using the shielding structure combined with 1 cm B4C/Al composites and 10 cm polyethylene,the neutron flux rate outside the tank to below 1×105 n/（cm2·s）can be reduced and the activation of surrounding equipment by neutrons can be prevented.Using the shielding combined with 1 cm B4C/Al composites and 20 cm polyethylene,the neutron dose rate outside the fuel salt tank can be reduced down to less than 2 m Sv/h,which meets the regulations for the safe transport of radioactive material.Chapter six,summary and outlook:The shielding performance of B4C/Al composites corresponding to different neutron energies has been summarized.The influence of B4C content,10B abundance,material density and B4C particle size to the neutron shielding performance of B4C/Al composites were summarized.Compared to traditional neutron shielding materials such as cadmium and concrete,B4C/Al composites as the neutron absorbing material of molten salt reactor fuel salt has neutron absorption ability in a wide energy range,which can reduce the space and weight of neutron shielding.The problems of material mechanical strength and radiation resistance need to be solved if the B4C/Al composites are used as structural neutron shielding materials in future. |
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