Abstract In this study, 12 different concentrations of shielding materials were developed and produced. They were covered with high temperature resistant (1500 °C) sodium silicate sealing paste. Epoxy resin was produced by adding different percentages of additive materials such as chromium oxide (Cr2O3), lithium (LiF), and nickel oxide (NiO). The GEANT4 and FLUKA codes of the Monte Carlo simulation toolkit were used to determine the mixing ratios. The total macroscopic cross-sections, effective removal cross-sections, mean free path, half value layer, and transmission neutron number were determined for fast neutron radiation using GEANT4 and FLUKA simulation codes. The mass attenuation coefficient, the effective atomic number and half-value layer (HVL) of the samples were calculated using Phy-X/PSD software. The absorbed dose was measured. In this study, an 241Am–Be neutron source with 74 GBq activity and average neutron energy of approximately 4.5 MeV and a BF3 gas detector were used. Both simulation and experimental measurements were compared with paraffin and conventional concrete. The new composite shielding material absorbed radiation much better than the reference materials. This new radiation shielding composite material can be used in nuclear medicine, transport and storage of radioactive waste, nuclear power plants, and as a shielding material for neutron and gamma radiation.

中文翻译：

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潜在的新型复合材料缓和中子伽马辐射的实验和蒙特卡罗模拟研究

摘要 本研究开发并生产了12种不同浓度的屏蔽材料。它们覆盖有耐高温（1500°C）的硅酸钠密封膏。环氧树脂是通过添加不同百分比的添加剂材料制成的，例如氧化铬 (Cr2O3)、锂 (LiF) 和氧化镍 (NiO)。Monte Carlo 仿真工具包的 GEANT4 和 FLUKA 代码用于确定混合比。使用 GEANT4 和 FLUKA 模拟代码确定了快速中子辐射的总宏观截面、有效去除截面、平均自由程、半值层和传输中子数。使用Phy-X/PSD软件计算样品的质量衰减系数、有效原子序数和半值层(HVL)。测量吸收剂量。在这项研究中，使用了具有 74 GBq 活度和大约 4.5 MeV 平均中子能量的 241Am-Be 中子源和一个 BF3 气体探测器。模拟和实验测量都与石蜡和常规混凝土进行了比较。新的复合屏蔽材料比参考材料更好地吸收辐射。这种新型辐射屏蔽复合材料可用于核医学、放射性废物的运输​​和储存、核电站，以及作为中子和伽马辐射的屏蔽材料。新的复合屏蔽材料比参考材料更好地吸收辐射。这种新型辐射屏蔽复合材料可用于核医学、放射性废物的运输​​和储存、核电站，以及作为中子和伽马辐射的屏蔽材料。新的复合屏蔽材料比参考材料更好地吸收辐射。这种新型辐射屏蔽复合材料可用于核医学、放射性废物的运输​​和储存、核电站，以及作为中子和伽马辐射的屏蔽材料。