Preparation of 99Mo from the 100Mo(γ, n) reaction and chemical separation of 99mTc

Optimization of target system for the production of 99Mo via 100Mo(γ,n)99Mo reaction

With an increase of stopping operation of nuclear reactors worldwide, the supply of medical 99Mo becomes difficult and thus many efforts have been made to find an alternative. A process based on an electron linear accelerator (linac) system and a100Mo target via the 100Mo (γ,n)99Mo reaction receives a lot of attention due to the relatively low level of co-produced impurities. This process has been recently developed at the Institute of Modern Physics (IMP) and the Monte Carlo simulation was used to optimize the target system before operating pilot irradiation experiments. First, tungsten and tantalum, as mostly used converter materials, were tested. The yield of 99Mo was evaluated with respect to the converter thickness and the electron beam energy by means of Geant4 simulations. Besides, the specific activity of 99Mo produced from one-stage approach (100Mo target without a converter) and two-stage approach (100Mo target with a converter) was compared when varying the testing conditions. The two-stage approach was selected for the experiment due to the higher specific activity of produced 99Mo at all tested conditions. A target consisting of a 10 mm thickness of the 100Mo tablets and a 2.4 mm thick Ta converter was irradiated for 40 h (50 MeV with 0.2 μA). The Geant4-calculated specific activity of generated 99Mo at the end of bombardment agreed well with the experimental value, which proved high level of accuracy of the Geant4 simulation. In future studies, the Geant4 simulation will be used to optimize the production process when using high power linac system.

Sponge-like Ca-alginate/Lix-84 beads for selective separation of Mo(VI) from some rare earth elements

In this investigation, a novel alginate complex was developed for the selective separation of molybdenum (Mo(VI)) ions from some rare earth elements (REEs). In this regard, alginate as a natural polysaccharide was impregnated and modified with 2-hydroxy-5-nonylacetophenone oxime (Lix-84) and characterized using FT-IR, TGA/DTA and SEM-EDX. The relation between medium acidity, adsorption kinetics, sorbent dose, isotherm models, temperature and Mo(VI) recovery was investigated. It was concluded that the impregnation stage promoted the Mo(VI) separation. The kinetics and isotherm data were well-fitted and matched with the pseudo-first-order model and Langmuir isotherm model; respectively. The Langmuir maximum adsorption capacity of Mo(VI) reached 72.2 mg/g. The developed material showed excellent separation performance towards Mo ions over the investigated REEs. The desorption and recovery of the loaded Mo(VI) ions were achieved using 1.0 M HCl. Reutilization of Alg/Lix-84 was confirmed up to three adsorption-desorption cycles with no damage of the beads as proved with SEM analysis. The adsorption mechanism of molybdenum onto Alg/Lix-84 was elucidated through FTIR and XPS measurements and was found to be governed by both electrostatic interaction and ion exchange. Therefore, the developed material has a promising potential for the selective separation of molybdenum from REEs-containing solution.

Development of LEU-based targets for radiopharmaceutical manufacturing: A review

⁹⁹Mo is an essential medical isotope that comprises of at least 70% of radioactive procedures globally. Currently an essential component of ⁹⁹Mo manufacturing is the uranium target from which ⁹⁹Mo is produced by fission. As the world moves towards low enriched uranium (LEU) targets due to non-proliferation concerns it is becoming of interest to find methods to increase the efficiency of the LEU targets in order to reduce the ever increasing nuclear waste levels of which a long term solution for disposal or treatment has yet to be satisfactorily found. Advantages and disadvantages of various target designs are investigated and discussed along current disposal and reprocessing methods. The idea of a reusable target is introduced as a way forward in reducing the nuclear waste burden for future generations.