| With the rapid development of nuclear electricity in China, the continuedsafe storage of spent fuel in the nuclear power plants is becoming a key issue inthe process of nuclear technology. However, the existing spent fuel storage withradiation shielding materials couldn’t meet the service requirement in structuralproperties, corrosion resistance and shielding performance at the same time. Inthis paper, the Monte Carlo method was applied to design a new kind ofaluminum-matrix which could meet the demand in structural and functionalintegration in the storage of spent fuel radiation shielding material. By taking the6061aluminum alloy as the carrier component, the addition content of boroncarbide as neutron absorbing component and lead as γ radiation shieldingcomponent is vital to provide a theoretical basis in material design.In this paper, MCNP5.0and Super MC/MCAM software was uesd to studythe relationship between the boron carbide content, reinforced partiledistribution, thickness of aluminum matrix composites and neutron transmissionratio, the interaction mechanism between neutron and aluminum boron carbidecomposite was discussed; the relationship between component content oflead-contained aluminum boron carbide composites, incident ray energy andneutron transmission ratio, secondary γ-rays absorption rate, radioactive γ-rayabsorption rate was studied. The impact of boric acid solution concentration andthe intensive design of spent fuel storage frame cells on radiation shieldingperformance was analyzed. The radiation shielding effect of aluminum matrixshielding materials under conditions of235U decay source generated from spentfuel was estimated. The results show that due to the presence of neutron absorption nuclides10B,B4C content has great influence on neutron absorption but limited impact on thesecondary γ-ray shielding performance. It is attributed to the small density ofB4C that the loss of material density may cause a γ-ray absorption decreasing;B4C particle size increases while neutron absorbing performance deteriorating;the neutron transport distance within the material,which was caused by B4Cneutron particle shape changing will affect the neutron absorption.Since the γ ray absorption nuclide Pb is of a large atomic number, massabsorption coefficient and a high density high γ rays, content of Pb will improvethe neutron shielding properties, and significantly affect both the secondaryγ-ray and decay γ-ray absorption capacity.The energy of the incident particles directly determines the radiationshielding effect of materials, which is due to the macroscopic absorption crosssection material generally decreases with increasing radiation energy. Thedecline tendency of neutron transmission coefficient and secondary γ-rayrelative absorption rate of increased gradually slows with the increasing ofincident neutron energy, the γ-ray absorption rate increases with the incidentγ-ray energy decreasing.Because the boric acid solution as spent fuel storage medium plays a vitalrole in moderating and then thermalizing the incident fast neutron, itsconcentration decreases with neutron absorption properties of aluminumshielding material increasing. The secondary γ-ray shielding performanceincrease with the concentration and spent fuel unit interval decreasing; Thechanges in the storage medium is not significantly affect absorbing of spent fuelgenerated decay γ rays. Under intensive storage of spent fuel assembliescondition (assembly interval as23cm), the shield thickness is of0.7cm,235Uspent fuel generated decay neutrons was fully moderating and thermalizingthrough a boron concentration of boric acid environment2500ppm. When the B4C/Al material with30wt.%B4C content was applied as the shield layer, it isobtained that98.04%of neutrons,45.44%of secondary γ rays and20.21%of aγ-ray was absorbed; When the Pb-B4C/Al material with30wt.%B4C contentand25wt.%Pb content was applied as the shield layer, it is obtained that98.82%of neutrons,61.05%of the secondary γ-rays and47.08%of γ-rays wasabsorbed. |
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