FAST AND THERMAL NEUTRON REMOVAL CROSS-SECTION FOR CERAMIC GLASS ALUMINUM OXYNITRIDE
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Date
2024-09-08
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Dergipark
Abstract
This study investigates the effectiveness of transparent aluminum oxynitride (AlON) in neutron shielding, focusing on both fast and thermal neutrons. Using conventional radiation attenuation parameters, the macroscopic neutron removal cross-sections of AlON were calculated for varying neutron energies and material thicknesses. The Geant4 simulation toolkit was employed to model and analyze the neutron interactions with AlON. The results indicate that AlON exhibits a high neutron shielding capacity for fast neutrons (2 MeV), with transmission factor values ranging from 0.783 to 0.260 for material thicknesses between 1 and 10 cm. These values are nearly identical to those for water, which range from 0.782 to 0.257, highlighting AlON's comparable performance. However, for thermal neutrons, AlON's performance was less effective, only surpassing lead but not concrete or water. The findings suggest that while AlON is highly effective for fast neutron shielding, it may require complementary materials to adequately shield thermal neutrons. This could involve using AlON in combination with other materials to create a more comprehensive neutron shielding solution. AlON shows significant potential as a neutron shielding material, particularly for fast neutrons. Its integration with additional shielding materials could enhance its overall effectiveness, making it suitable for various nuclear and radiation protection applications.
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Keywords
Neutron, Attenuation, Shielding, Geant4, Cross-section