Volume 113, Issue 3

In preparation of gas-phase chemical experiments with moscovium (Mc, element 115), we studied the chemical behavior of the short-lived bismuth radioisotope 211 Bi in helium, argon, and oxygen atmosphere. For that purpose, we performed off-line isothermal gas chromatography experiments at room temperature. Using different carrier gases, the short-lived volatile 219 Rn precursor, provided from an 227 Ac-source, was transported through the R ecoil T ransfer C hamber (RTC) at the gas-filled separator TASCA and into the mini - C ryo- O nline M ulti detector for P hysics a nd C hemistry of T ransactinides (miniCOMPACT) chromatography and detection setup. Internal chromatograms were recorded as a function of various parameters including carrier gas type and flow rate, thus characterizing the novel miniCOMPACT detector array. This aids to optimize the conditions for experiments with superheavy elements. The bismuth progeny of 219 Rn deposited on the SiO 2 surface of the miniCOMPACT via diffusion-controlled deposition. Bismuth showed the expected high reactivity towards the SiO 2 surface of the miniCOMPACT. Experiments in argon and oxygen atmosphere showed no measurable differences in the deposition distribution of the activity. The intermediate 36-min 211 Pb, a member of the 227 Ac decay chain feeding the studied bismuth isotope, was taken into account. To extract thermodynamical data from the results, namely the lower limit of the value of the adsorption enthalpy (−Δ H ads ) of Bi on SiO 2 , we performed Monte Carlo simulations, adapted to account for the precursor effect, and compared the experimental results to their output. Simulations were also performed for bismuth’s heavier homologue, moscovium, using a theoretically predicted value for −Δ H ads of this element on SiO 2 . These suggest moscovium to adsorb in the first part of the miniCOMPACT detector array, in line with recent observations.

The uptake of Eu, Th, U, and Pu by Korean granite and biotite gneiss was investigated in a series of batch experiments. Experiments were conducted under well-defined redox conditions, i.e. , oxidizing (air), mildly reducing (Ar-atmosphere and buffered with hydroquinone, pe + pH ≈ 8–9), and strongly reducing (Ar-atmosphere and buffered with Na 2 S 2 O 4 , pe + pH ≈ 0.5–3). Radionuclide concentration, pH and E h were systematically monitored up to t  ≤ 113 days after the addition of the radionuclide. The natural content of Eu, Th and U in pristine granite and biotite gneiss materials was quantified by means of alkaline fusion. Eu exhibited moderate sorption on biotite gneiss displaying higher distribution ratios ( R d ) compared to granite. This observation was possibly explained by the affinity of Ln(III)/An(III) towards biotite mineral absent in the investigated granite material. Strong sorption was observed for Th, U, and Pu in reducing systems where the predominance of the +IV oxidation state is expected. For these three systems, the strength of the uptake follows the order R d (Pu(IV)) >  R d (U(IV)) >  R d (Th(IV)), consistent with the hydrolysis strength of the corresponding aquo-ions. A significantly weaker sorption was observed for U and Pu under oxidizing conditions, although R d values are manifestly higher for Pu than U. Thermodynamic calculations for the oxidizing conditions predicted the predominance of U(VI) and Pu(V)/Pu(IV), explaining observed differences in retention under oxidizing conditions. These results contribute to a quantitative description and a better understanding of the retention of redox-sensitive radionuclides in crystalline host rocks. Emphasis is placed on the importance of utilizing both redox-stable probes ( e.g. , Eu, Th) and redox-sensitive actinides ( e.g. , U, Pu), as well as well-defined redox conditions for accurate predictions of radionuclide retention.

The aim of this research is to present a new, fast, simple, low-cost and innovative method for the production and purification of pharmaceutical grade indium-111 chloride for use in nuclear medicine.In the previous methods for cyclotron production of indium-111 radioisotope, besides the need for high amount of enriched cadmium, in electrodeposition of enriched cadmium on copper backings for target preparation and its chemical dissolution step, impurities of isotopes of copper and zinc enter the final product. So, radiolabeling of such products was hard or impossible with some molecules. In this research, due to the proper selection of the target isotope, enriched cadmium-112, and the optimal proton bombardment energy, only very low amount radioisotopic impurities were produced and observed in the sample solution. After the separation process, the amounts of zinc-65 and indium-114 were less than the standard permissible amounts in indium-111 radiopharmaceuticals. The concentration of cadmium impurities in the final product was less than 0.1 ppm. The radionuclidic purity of the sample was confirmed by gamma ray spectroscopy, and it was found to be suitable for use in preclinical studies.

99m Tc, the daughter product of 99 Mo, is a γ-emitting radionuclide with essential diagnostic applications in nuclear medicine. In this paper, an accelerator-based method ( 100 Mo (γ, n) 99 Mo) for production of 99 Mo and 99m Tc has been explored. Approximately 68.3 MBq of 99 Mo was successfully produced by the irradiation of 100 Mo metallic target for 40 h using electron beam with energy of 50 MeV and current of 0.2 μA at electron linear accelerator (Institute of Modern Physics, Chinese Academy of Sciences, IMPCAS). Different types of 99 Mo/ 99m Tc generators were prepared using hierarchical macro/mesoporous γ-Al 2 O 3 (HMMA) as column adsorbent, and their performances were evaluated for over one week. 99 Mo/ 99m Tc generator with a column (3 mL, 4.2 × 0.95 cm) packed with 0.6 g HMMA exhibited excellent eluting performance. 99m Tc could be collected within 4.0 mL of saline solution with high purity, and the elution efficiency could reach >85 %. Furthermore, 99 Mo breakthrough in the eluates was negligible (<2 × 10 −3  %) and concentration of impurity (<10 ppm) was acceptable. Finally, the enriched 100 Mo was eluted from the spent 99 Mo/ 99m Tc generator using 1.0 mol/L ammonium hydroxide, and then reduced by high-temperature hydrogen reduction process with a total recovery of 95.3 %. This work demonstrated that the preparation of 99 Mo via 100 Mo (γ, n) 99 Mo reaction and isolating 99m Tc using HMMA column chromatography have a potential in application.

Male prostate cancer (PCa) is considered among the most fatal illnesses. Despite the recent decrease in prostate cancer incidence attributed to advancements in early detection and therapy, these reductions have not effectively mitigated the elevated fatality rate linked to this disease. The drug Nilutamide was effectively radiolabeled with technetium-99m, producing a radiochemical yield of 96 ± 0.14 % under optimal conditions. In our study, two cohorts of mice were utilized, namely the control group and the group with prostate cancer. Various biochemical parameters, including PSA levels in serum, were assessed, revealing a significantly elevated value in the group with prostate cancer, indicating potential tumor development. Furthermore, the activities of antioxidant enzymes (CAT, SOD) were notably lower in the group with prostate cancer compared to the healthy control group, while the oxidative activity reflected by MDA levels, the final product of lipid peroxidation, was higher in the prostate cancer group than in the healthy control group. The biodistribution analysis showed rapid localization of 99m Tc-Nilutamide in prostate cancer tissue after 2 h post-injection, with a substantial value of 11.4 ± 1.1 % I. D/g tissue. Consequently, it was deduced that radiolabeled 99m Tc-Nilutamide can serve as an effective imaging tool for prostate cancer.

The authors present here the decomposition of un-irradiated (pristine) as well as of gamma ( γ ) and electron beam (EB) irradiated samples of europium (III) acetate hydrate (EuAc.xH 2 O) in the temperature range of 25–900 °C in the air atmosphere. Two absorbed doses of 10 3 ( γ -ray) and 10 2  kGy (EB) were examined. The profiles of the TG curves of the dehydration process display noticeable changes in induction periods and mass loss percentages by exposure to irradiation. The kinetics of the dehydration process were analyzed using both model-fitting and model-free approaches. The dehydration process was controlled by the phase boundary model (R 2 ). The E a  −α plots indicate that the dehydration is not a complex process and follows one reaction mechanism. Powder X-ray diffraction displayed that europium acetate hydrate crystallizes in a monoclinic system (SG P 2/ m) , and no phase transformation was detected by two sources of irradiation up to 10 3 ( γ -ray) and 10 2  kGy (EB). Thermodynamic parameters of the dehydration process were calculated and assessed. A predicted thermogram (TG) of the isothermal dehydration of EuAc.xH 2 O was constructed from non-isothermal data and used to determine the reaction model and the kinetic parameters of the dehydration process.

Organoarsenic can be applied to feed to improve feed efficiency or as a radiation dosimeter, as presented in this study. A developed radiochromic solution and powder based on p -arsanilic acid (P-ASA) was evaluated at various concentrations. Using a UV–vis Spectrophotometer, EPR analysis were operated, the spectrophotometric characteristics of P-ASA aqueous solutions that were both unirradiated and gamma-irradiated were evaluated. For applications involving moderate dose dosimetry, two absorption peak intensities were observed at λ max 300 nm and 361 nm following gamma ray exposure within the range of up to 4.5 kGy. The results indicate that P-ASA solutions exhibit good dosimetric properties and can be considered as a dosimeter.

The current study investigates the presence 222 Rn in tap water, natural spring water, and well water from various locations in Er-Rachidia, Morocco, to assess the potential radiological risks posed to the local community. Radon, a radioactive gas produced by the decay of uranium and thorium in rocks and soil, poses significant radiation hazards. Using a Nuclear Track Detector (LR-115), researchers investigated the presence of radon in water from wells, dams, and rivers. The highest radon concentrations are found in well water, with an average of 1.92 Bq/L. The concentration of river water is the lowest, coming in at 0.53 Bq/L, while the average concentration of dam water is 0.84 Bq/L. On average, the annual effective dosage for well water is 4.57 ms per year. The water dosages from dams range from 1.74 to 2.44 μSv/y, with 1.98 being the average. River water ranges from 0.51 to 1.69 mSv/y, averaging 1.24 µSv/y. The excess lifetime cancer risk varies significantly across water sources: 1.27 × 10−3 to 1.84 × 10−3 (average 1.60 × 10−3) for well water, 0.61 × 10−3 to 0.85 × 10−3 (average 0.69 × 10−3) for dam water, and 0.18 × 10−3 to 0.59 × 10−3 (average 0.43 × 10−3) for river water. The results dispel fears of significant radioactive dangers to the locals by suggesting that the measured radon concentrations are within the limits set by international organizations.