A new68Ge/68Ga generator using CeO2 as absorbent for68Ge has been developed. Sharp elution curves were obtained by using 0.02 mol/l HCl as eluent. About 56% of68Ga formed on the column was concentrated in 0.5 ml of eluate. Neither radioactive impurities, nor dissolved cerium were found in the eluate.
Authors:Cao Benhong, Li Zongquan, and Wang Yongxian
The absorption behavior of68Ga(III) and68Ge(IV) on three different type of ferric oxides has been studied. Alpha-ferric oxide in trigonal structure is found to be
a suitable material for the generator. The elution yield of68Ga from the column of trigonal crystal of α-ferric oxide with absorbed68Ge(IV) can reach 90% in 1 ml 0.05 mol/l HCl solution, while the breakthrough of68Ge was less than 1.8·10−4. This generator is stable and resistant to chemicals and radiation. The content of Fe in 1 ml eluate is less than 0.1 ppm.
The recovery of68Ga was higher and more constant than that of the known68Ge−68Ga generator system.
Authors:Wu-Long Cheng, Yun Jao, Chung-Shin Lee, and Ai-Ren Lo
A novel type of a binary Ga/Ag alloy electrodeposited on silver substrate as a solid target was developed. It was successfully used for the preparation of 68Gc/68Ga generator. The deposition was carried out in an alkali solution containing gallium, silver and certain electrolytes at controlled current and ambient temperature so that the quality of the deposits was proved to be suited for target irradiation. The yield of 68Ge with proton bombardment on this deposits via the 69Ga(p,2n) reaction was assessed. The chemical process for providing millicuries 68Ge/68Ga generator using a generic tin dioxide as an adsorbent was also established. It was revealed from long-term elution tests that approximately 60–70% of 68Ga could be eluted from the generator column with 4 ml of 1.0M HCl per elution, and high radio- and chemical purities of the eluates were quite satisfactory for application purposes.
Authors:M. Shehata, B. Scholten, I. Spahn, S. Qaim, and H. Coenen
The radiochemical separation of radiogallium from radiogermanium was studied using ion-exchange chromatography (Amberlite
IR-120) and solvent extraction (Aliquat 336 in o-xylene). Both Amberlite IR-120 and Aliquat 336 in o-xylene have been used for the first time in separations involving radiogallium and radiogermanium. For tracer studies the
radionuclides 68Ge (t1/2 = 270.8 days), 69Ge (t1/2 = 39 h) and 67Ga (t1/2 = 78.3 h) were used. They were produced by the nuclear reactions natGa(p,xn)68,69Ge and natZn(p,xn)67Ga, respectively, and separated from their target materials in no-carrier-added form. Several factors affecting the separation
of radiogallium from radiogermanium were studied and for each procedure the optimum conditions were determined. The solvent
extraction using Aliquat 336 was found to be better. The separation yield of radiogallium was >95%, the time of separation
short, the contamination from radiogermanium <0.008% and the final product was obtained in 0.5 M KOH. This method was adapted
to the separation of n.c.a. 68Ga from its parent n.c.a. 68Ge. The quality of the product thus obtained is discussed.
The method described in this paper is a new and more capable separation (5 ppm zinc impurity) as well as fast with a 25–35 min
whole process time. Optimal 67Ga separations (yielding 93.2% efficiency) from Cu and 68Zn were obtained by precipitate with 2 M NaOH.
Systematic investigations into the practical problems of labelling transferrin with68Gain vitro are presented. A chemical purification of the68Ga generator eluate is described and the working conditions whereby several millicuries of68Ga may be bound quantitatively onto a few milligrams of transferin are defined.
Authors:Erik de Blois, Ho Chan, Kamalika Roy, Eric Krenning, and Wouter Breeman
PET with 68Ga from the TiO2- or SnO2- based 68Ge/68Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity (68Ge vs. 68Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of 68Ge activity is produced by eluting the 68Ge/68Ga generators and residues from PET chemistry. Since clearance level of 68Ge activity in waste may not exceed 10 Bq/g, as stated by European Directive 96/29/EURATOM, our purpose was to reduce 68Ge activity in solution from >10 kBq/g to <10 Bq/g; which implies the solution can be discarded as regular waste. Most efficient
method to reduce the 68Ge activity is by sorption of TiO2 or Fe2O3 and subsequent centrifugation. The required 10 Bq per mL level of 68Ge activity in waste was reached by Fe2O3 logarithmically, whereas with TiO2 asymptotically. The procedure with Fe2O3 eliminates ≥90% of the 68Ge activity per treatment. Eventually, to simplify the processing a recirculation system was used to investigate 68Ge activity sorption on TiO2, Fe2O3 or Zeolite. Zeolite was introduced for its high sorption at low pH, therefore 68Ge activity containing waste could directly be used without further interventions. 68Ge activity containing liquid waste at different HCl concentrations (0.05–1.0 M HCl), was recirculated at 1 mL/min. With Zeolite
in the recirculation system, 68Ge activity showed highest sorption.
Authors:T. Katabuchi, S. Watanabe, N. Ishioka, Y. Iida, H. Hanaoka, K. Endo, and S. Matsuhashi
The radionuclide 67Cu was produced via the 68Zn(p,2p)67Cu reaction by irradiating enriched 68Zn targets with 70 MeV proton beam. Copper-67 was chemically separated from the zinc target by ion-exchange chromatography
using Chelex-100 chelating ion-exchange resin. Procedure for recovery of the enriched 68Zn was developed. The target recovery yield of this method was evaluated to be more than 97%.
Authors:S. Egamediev, S. Khujaev, and A. Mamatkazina
The influence of preliminary annealing (at 400 and 1000 °C) and wet treatment (with 0.1M HCl; with 0.1M NaOH) of aluminum oxide on the separation efficiency of 68Ge-68Ga radionuclide chain was studied. The adsorption behavior of 68Ga daughter radionuclide was examined by desorption in order to find the best conditions for separation of both radionuclides. The effect of the preliminary annealing and wet treatment of alumina on the experimental generator columns was studied.