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131I is an important radioisotope in nuclear medicine, but due to the inherent defects of wet separation technology，the separation of 131I from uranium fission products (UFPs) has serious environmental pollution and low extraction efficiency. Considering the high yield of 131I in UFPs, it is necessary to develop a new method for separating 131I from UFPs, because it can not only expand the ways to obtain 131I, but also improve the utilization efficiency of UFPs. Unlike the traditional wet chemical separation technique, 131I was separated from UFPs by dry distillation in present work. Three stages worked under different temperatures, such as powdered the target, dry distillation and heat preservation, were involved in the separation process, so as to acquire as much as 131I from the target. The effect of dry distillation temperature on 131I separation efficiency was systematically investigated, it was found that the separation efficiency of 131I was higher than 98% when the distillation temperature was over 950 ℃. Moreover, it was also observed that both radioiodine and 103Ru were sublimated from the UFPs at the optimum distillation condition, but only radioiodine existed in the product. In order to explain this phenomenon, a detailed analysis of the iodine separation process was carried out. By combining the experimental results with the theoretical calculations, the separation mechanism of radioiodine from 103Ru was estimated as: 103Ru volatilized from the UFPs through the formation of RuO4, the volatile compound of 103Ru may be decomposed into RuO2 during the migration because of the high temperature in the heating tube, and then settled down. Therefore, to separate 131I from UFPs by dry distillation and obtain 131I product with high radiochemical purity, it is necessary to rationally plan the separation process, at the same time, the scientifically design of the heating tube length is also meaningful.

The aim of present study was to determine the efficiency of commercially available PEG-based resins for separation of 99mTc from irradiated 100Mo target. Stable and radioactive tracer studies demonstrated that these PEG-based resins have extreme selectivity toward ReO4− (99mTcO4−) over MoO42− ions. Moreover, PEG-based resins still have 95% adsorption efficiency even after 600 kGy γ-irradiation, suggesting the excellent irradiation stability. The adsorption rate of ReO4− and 99mTcO4− rapidly increased to 90% in 10 s. Subsequently, three-column chromatographic processing based on these superior resins for isolating 99mTc was developed and examined. The separation investigation of 99mTc medical radioisotope produced by 100Mo(p, 2n)99mTc reaction was successfully performed. The recovery yield and the radionuclidic purity of 99mTc was about 90% and 99.9%, respectively. Moreover, the separation of 99mTc from a low specific activity 99Mo was also investigated. We are excited about the potential of this procedure with these accessible commercial availability PEG-based resins to address 99mTc separation problems and increase access to medical nuclide 99mTc for the general community. [ABSTRACT FROM AUTHOR]

Uranium tetraboride (UB4) was successfully synthesized by boro/carbothermal reduction of UO2 with B4C and carbon as combined reduction agents under flowing argon. The effects of processing temperature and mole ratios of starting materials on phase evolution were studied. XRD results demonstrated that UB4 was obtained with 3.75 mol% B4C in excess based on the UO2/B4C/C molar ratios of 1:1:1 at 1500°C. SEM observation revealed that submicrometer‐sized quasi‐spherical UB4 particles cemented together to be an aggregate. The laser particle size analysis showed that the particle size was in the range of 1‐10 μm. The oxidation behavior of UB4 was also investigated by TG and XRD. The oxidation of UB4 started at about 500°C and it showed better oxidation resistance than other basic uranium nuclear fuels (UO2, UC, UN and U3Si2). The oxidation chemical process of UB4 was presented as a three‐step process: (a) the formation of U3O8 and B2O3 oxidation products; (b) the formation of UB2O6 intermediate product by the interaction of U3O8 and B2O3; (c) the decomposition of UB2O6 to get U3O8. [ABSTRACT FROM AUTHOR]

99m Tc is regarded as the most important medical isotope, and its supply issues have garnered significant attention. A simple and efficient separation method was performed for the production of 99m Tc from 100 Mo target in this study. The entire process involves accelerator irradiation, 99m Tc/ 100 Mo separation, and target material recovery. The key aspect is separation process, which including the high-temperature conversion of metal molybdenum targets and the selective solution of 99m Tc with normal saline. This method can separate highly pure 99m Tc within 1.5 h, with a separation efficiency exceeding 80%. The reagents used in the separation process are minimal, resulting in less radioactive waste. Additionally, the target material is easy to reclaim, with a recovery rate of over 95%., Competing Interests: Declaration of competing interest The authors declared no potential conflicts of interest with respect to the research, author-ship, and/or publication of this article., (Copyright © 2024 Elsevier Ltd. All rights reserved.)

Hierarchically macro-/mesoporous γ-Al 2 O 3 (HMMA) was synthesized and characterized by various analytical techniques. The results indicated that HMMA possessed macropores (∼0.45 μm) and mesopores (∼10.6 nm) with a large surface area (∼542 m 2  g -1 ). The absorption behaviors of Mo and Re with HMMA were investigated. The maximum static absorption capacity could reach about 250 mg Mo per g HMMA. The absorption equilibrium can be attained quickly within 10 mins. At initial Mo ions concertation of 10,000 mg L -1 , the breakthrough capacity was determined to be around 200 mg Mo per g HMMA. Additional, absorption mechanism results indicated that Mo ions reacts strongly with a hydroxyl on the surface of γ-Al 2 O 3 and an adjacent Al atom, simultaneously. A 9.15 mCi (339 MBq) 99 Mo generator was prepared and evaluated its performance for over one week. The recovery of 99m Tc could reach about 89% with favorable radionuclidic, radiochemical and chemical purity for nuclear medicine application. HMMA has a potential application prospect for the preparation of low specific activity (LSA) 99 Mo/ 99m Tc generator., (Copyright © 2021 Elsevier Ltd. All rights reserved.)