A comprehensive study of the shielding ability from ionizing radiation of different mortars using iron filings and bismuth oxide.

The current work discusses the radiation attenuation capability and different shielding characteristics of different mortar samples. The samples were prepared by replacing different percentages of fine aggregate with iron filling and replacing different percentages of hydrated lime with Bi2O3 (0-50 wt.%). The prepared mortar samples are coded as CHBFX where X = 0, 10, 30, and 50 wt.%. The mass and linear attenuation coefficient was determined experimentally using a narrow beam technique, where a high purity germanium detector, and different point gamma-ray sources (such as Am-241, Cs-137, and Co-60). The linear attenuation coefficient was also calculated using the Monte-Carlo simulation code and the online Phy-X/PSD software. The comparison of the three methods showed a good agreement in the results. The linear attenuation coefficient drops from 19.821 to 0.053 cm-1 for CHBF0, from 27.496 to 0.057 cm-1 for CHBF10, from 42.351 to 0.064 cm-1 for CHBF30, and from 55.068 to 0.071 cm-1 for CHBF50 at photon energy range from 0.015 to 15 MeV. The half-value layer thickness, tenth-value layer thickness, and mean free path of the prepared mortar composites were also calculated photon energy ranged from 0.015 to 15 MeV. The fast neutron removal cross-section of the prepared CHBFX mortar samples have values of 0.096 cm-1, 0.098 cm-1, 0.103 cm-1, and 0.107 cm-1 for the mortar samples CHBF0, CHBF10, CHBF30, and CHBF50, respectively. The results showed that the mortar sample with the highest iron filing concentration, CHBF50, provides the best protection against gamma rays and fast neutrons which could be used in the nuclear and medical fields.

Radionuclides distribution and radiation hazards assessment of black sand separation plant's minerals: a case study.

This study assessed the radioactivity levels and associated risks in the black sand-separated products obtained from the black sand separation plant in Delta, Egypt. A total of sixteen samples were taken from hot spots during and after the separation process. These include water samples and other samples that represent monazite, rutile, zircon, granite, ilmenite, and silica products. The hot spots included the area where the ore was stored. The activity concentrations of 232 T h , 226 R a , and 40 K were determined in these samples using a p-type HPGe detector. Based on gamma spectrometric analysis, samples of rutile, zircon, and monazite had the highest amounts of radioactivity because they contained the highest NORM's activity concentrations. In addition, it indicated that the radiological hazard indices of the collected samples were higher than the average world limits for sand texture. These findings suggest that the black sand separation process reveals potential risks to human health and the environment, and therefore, appropriate measures need to be taken to mitigate these risks, especially for the safety of the workers on-site. Reducing the risk associated with those sites should be controlled by implementing the recommendations declared for the series of International Basic Safety Standards of the International Atomic Energy Agency (GSR) Part 3, as affirmed in Document No. 103 of 2007 by the International Commission on Radiological Protection (ICRP) as will be presented in the paper body.

Experimental, analytical, and simulation studies of modified concrete mix for radiation shielding in a mixed radiation field.

The current study assessed two concrete mixes prepared using dolomite and barite/limonite aggregates to shield against both energetic photons and neutrons. After that, a designed mix which comprised barite/goethite aggregates plus fine-powdered boron carbide additive, was proposed to improve the overall radiation shielding properties and in the same time, doesn't compromise or even improve the physic-mechanical properties of the mature concrete. The assessment started first with intensive experimental investigations to investigate the prepared mixes' shielding capabilities against both γ-rays and fast neutrons. Then, analytical computations were performed via number of reliable software programs such as; Phy-X, NXCom, MRCsC, JANIS-4, and MCNP5, in order to confirm the experimental results and to validate the created Monte-Carlo models. Finally, an intensive radiation shielding assessment for all concrete mixes understudy using, mainly, the validated MCNP models, was performed. The obtained results have revealed the superiority of barite mixes over the dolomite mix concerning attenuating photons moreover, the proposed designed mix has shown superiority over the other two prepared mixes considering shielding against; energetic photons, fast/thermal neutrons, and secondary emitted γ-rays, which nominates this mix to be a suitable universal shield that can be used even in mixed radiation fields.

A semi-empirical method for efficiency calibration of an HPGe detector against different sample densities.

In this work, a semi-empirical equation in terms of γ-energy, and sample density is derived, proposed, benchmarked, and applied for the peak efficiency calibration of an HPGe detector with respect to an axial source-to-detector configuration. The samples are in the form of cone-shaped Marinelli beakers of different densities in the range 0.7-1.6 g/cm3. The method employs the experimental measurements with the ANGLE-3 software calculations using the efficiency transfer method. The peak efficiency curve of an HPGe detector is calculated using the experimental measurements of point-like sources (133Ba, 137Cs, and 60Co). The ANGLE-3 software is then used to calculate the peak efficiency curves for samples with different densities in the γ-energy range 81-1332 keV. The peak efficiency curves are then fitted to get the energy coefficient; in addition, a linear relationship is then constructed between the energy coefficients and sample densities to get the density coefficients, and the derived equation as well. The derived equations are benchmarked using the peak efficiency curves by ANGLE-3 software in comparison with that the equation results. The results are found to be in agreement with an average relative error of about 1.5%. In addition, the derived equations are applied to estimate the activity concentration of radionuclides present in 5 cone-shaped samples with different densities using experimental measurements. The activity results are found to be in agreement with the certified values with an average relative error of about 2%. The limitation of the proposed equations is also discussed with respect to different material densities and different chemical compositions and correction factors for material composition self-attenuation for various materials are also presented.