Volume 112, Issue 1

Effective treatment of uranium-containing wastewater is of great significance to the sustainable development of nuclear power and the protection of ecological environment. In this study, a highly efficient uranium adsorbent, graphene oxide (GO)/nano-hydroxyapatite (nHA) composite microspheres (nHA@rGO) was synthesized, which could effectively remove uranium from aqueous solution. Under the condition of pH = 3.5, T  = 298 K, the maximum adsorption capacity reached 1672.96 mg/g. The results of batch experiments showed that the adsorption capacity of nHA@rGO microspheres was higher than that of nHA microspheres, indicating the enhancement of GO. The adsorption kinetics conformed to the pseudo second-order model. The changes of nHA@rGO microspheres before and after uranium adsorption were analyzed by FT-IR, XPS and XRD. The mechanisms of U(VI) ions adsorption onto nHA@rGO microspheres involved precipitation, surface complexation and ion exchange, in which the hydroxyl and phosphoric acid groups played important roles. The results showed that the prepared nHA@rGO microspheres can be used as an efficient and promising adsorbent for the treatment of uranium-containing wastewater.

Using batch and column procedures, the present work investigated the sorption behaviour of 137 Cs and 152+154 Eu by bentonite phosphate modified with nickel (BPN) sorbent. The kinetic data obey pseudo-1 st -order for 137 Cs and follow pseudo-2 nd -order for 152+154 Eu. Various sorption isotherm models were used to analyze equilibrium data. The thermodynamic functions reflect an endothermic and spontaneous sorption process. HCl (about 95.35 %) and CaCl 2 (about 98.13 %) showed the optimum eluents for the complete recovery of both 137 Cs and 152+154 Eu, respectively. Finally, column data show that 137 Cs and 152+154 Eu may be loaded on BNP sorbent and separated from an aqueous solution using a variety of HCl concentrations as eluent. The obtianed results revealed that BNP sorbent is suitable for recovering 137 Cs and 152+154 Eu from low-level radioactive waste effluents (LLW).

This work is interested in the sorption and separation of 131 Ba, 109 Cd, 152+154 Eu, and 97 Zr from radioactive solutions onto barium molybdenum titanate loaded on carboxy methyl cellulose (BaMoTi@CMC) composites. In this work, different samples of BaMoTi@CMC composites were fabricated by the co-precipitation method and characterized using different analytical tools such as X-ray diffraction (XRD), attenuated total reflectance (ATR), and scanning electron microscope (SEM). The batch sorption investigations on 131 Ba, 109 Cd, 152+154 Eu, and 97 Zr include the influence of time, pH, and metal ion concentrations. The data reveal that S-3 has higher sorption efficiency than S-2 under all conditions. Isotherm is studied by Langmuir and Freundlich models. Binary systems data confirm that Cd( ii ), Ba( ii ), and Zr( iv ) can be separated from Cd–Eu, Ba–Eu, and Zr–Eu binary systems using S-2 and S-3 at different pHs. Finally, the data prove that Zr( iv ) and Ba( ii ) can be easily separated from tertiary systems (Zr–Ba–Cd) onto S-2 and S-3 at pH 2.

Rhodium-103m is one of the most attractive Auger electron emitters for internal radiotherapy. The half-life of 103m Rh is relatively short (56.114 min). Therefore, it needs to be produced using a generator for clinical use. Most studies of 103 Pd/ 103m Rh generators using anion-exchange resins were carried out over 50 years ago. However, these resins are no longer commercially available. In the present study, we tested a 103 Pd/ 103m Rh generator using alternative anion-exchange resins (i.e., IRA904, IRA410, SA20A, and SA11AL). No-carrier-added 103 Pd was used to make the generators. The 103m Rh product was eluted from the generators using 6 mL of 0.1 M HCl with a flow rate 0.5 mL/min. The generator made from SA11AL showed good performance, with a yield of 39 %, an impurity level of 103 Pd in the product of 0.29 %, and an operation time of 14 min. This makes this generator competitive with previously developed ones.

This work focused the determination of the rare earth elements (REE) in Algerian bentonites named Mos and Mag using k 0 -NAA method. Eleven rare earth elements such as Ce, Nd, La, Sc, Sm, Yb, Tm, Ho, Tb, Eu and Lu were determined. Results revealed that the total light REE (LREE) in Mos (168 ppm) is higher than in Mag (68.5 ppm). In addition, the ratio of total LREE (LREE) to the total heavy rare earth elements (HREE) found to be 6.40 in Mag and 27.6 in Mos which indicates that Mos is highly rich with REE than Mag. Ce and Eu correspond to the highest and lower trace of REE in both bentonites, respectively. The concentration of Sc, Tb and Ho are comparable in two materials. All results are compared with other data of literature in terms of La, Sc, Ce, Nd, Sm and Yb. Amounts and the proportion of LREE comparing to THREE are discussed in this study.

The aim of this study is to create a radiological map of Kırşehir province in Turkey by analyzing the radioactivity levels of soil samples and environmental radiation measurements. 226 Ra, 232 Th, 40 K and 137 Cs radionuclides were analysed in soil samples collected from 47 locations using gamma spectrometry with an HPGe detector. According to the results obtained, gamma radioactivity concentration ranges and mean values for 226 Ra, 232 Th, 40 K and 137 Cs were determined as 133 ± 15–1515 ± 128 Bq/kg (599.40 Bq/kg), 0.3 ± 0.09–21.1 ± 1.7 Bq/kg (4.61 Bq/kg), 8.7 ± 0.9–128.5 ± 8.5 Bq/kg (29.66 Bq/kg) and 11.6 ± 8.6–273.8 ± 19.9 Bq/kg (48.80 Bq/kg) respectively. Average radioactivity levels for 40 K and 232 Th are above the world average. The mean radium equivalent activity (Ra eq ), gamma dose rate ( D ), annual effective dose equivalent (AEDE), external hazard index ( H ex ), and the excess lifetime cancer risk (ELCR) were calculated as 145.45 Bq/kg, 122.13 nGy/h, 0.15 mSv/h, 0.39 and 0.29 × 10 −3 respectively.

In this contribution, the distribution of naturally occurring radionuclides in the area around Main Karakoram Thrust (MKT) in Karakoram Range, North Pakistan is documented. Three natural radionuclides ( 226 Ra, 232 Th, and 40 K) and one anthropogenic radionuclide ( 137 Cs) were studied for their specific activities in 30 samples. The measurements were made by high resolution gamma-ray spectrometry. The sampling area is located in Gilgit Baltistan province of Pakistan at an altitude of 1838 m/6030 ft above sea level. MKT separates the Karakoram plate from the Kohistan-Ladakh Terranes and Indian Plate to the south. The specific activity varied as 4.5–56.5 Bq kg −1 , 18.2–61.4 Bq kg −1 , 1.4–19.6 Bq kg −1 and 51–1640 Bq kg −1 for 226 Ra, 232 Th, 137 Cs and 40 K, respectively. The average radium equivalent activity was 127.8 ± 45.9 Bq kg −1 . The external hazard index was <1 for all samples and representative level index was <1 for majority of the samples. The average air absorbed dose rate was 60.9 ± 23.2 nGy h −1 corresponding to the outdoor effective dose rate of 73.7 ± 28.0 μSv y −1 . These values were slightly higher than the world average values for air absorbed dose rate (51 nGy h −1 ) and outdoor annual effective dose rate (70 μSv y −1 ). The data revealed significant positive correlation between 226 Ra and 40 K. Principal component analysis revealed distribution patterns within the samples and identified three distinct groups. Data was also evaluated for the concentrations of uranium, thorium and potassium and their ratios.