Effect of the B4C content on microstructure, microhardness, corrosion, and neutron shielding properties of Al-B4C composites

Aluminum-Boron Carbide (Al-B4C) metal matrix composites (MMCs) are frequently studied because of their high thermal conductivity and ability to absorb neutrons, as an alternative material for the manufacture of baskets in spent fuel casks in the nuclear industry. Keeping in mind, this study aimed to produce Al-B4C MMCs and examine their properties to shield against neutron radiation. The Al-xB4C (x = 5, 10, 15, 20, 25, 30, 40, 45, and 50 wt%) composite powders were prepared by a high-energy planetary ball milling. The prepared powders were com-pacted with uniaxial cold compaction at 500 MPa and sintered in a tube furnace at 600 degrees C for 1 h under an Argon atmosphere. The microstructure of the obtained MMCs was characterized by using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). The XRD result indicates that the synthesized Al-B4C MMCs have a predominant phase of Al and B4C. The SEM images show the distri-bution of B4C reinforcement inside the Al matrix. The relative density, measured by the Archimedes principle decreased from 99.95 to 91.55% as the B4C ratio increased. The microhardness value was increased in the beginning from 44 to 75.5 HV and then decreased to 36.5 HV as the B4C ratio increased. The polarization curves show that the corrosion current density and the corrosion rate were significantly increased from 52.25 to 375.36 mu A cm-2 and 0.573-4.121 mm/year as the B4C ratio increased. The simulation results, computed by the MCNP6.2 program show that the thermal neutron cross-section and the fast neutron cross-section was increased from 4.5 to 40.1 cm-1 and 0.13 to 0.15 cm-1 as the B4C ratio increased. Further, the neutron equivalent dose rate was calculated experimentally with the 241Am-Be neutron source and the data was completely agreed with the computational results.