Methods of increasing the performance of radionuclide generators used in nuclear medicine: daughter nuclide build-up optimisation, elution-purification-concentration integration, and effective control of radionuclidic purity.

Methods of increasing the performance of radionuclide generators used in nuclear medicine radiotherapy and SPECT/PET imaging were developed and detailed for 99Mo/99mTc and 68Ge/68Ga radionuclide generators as the cases. Optimisation methods of the daughter nuclide build-up versus stand-by time and/or specific activity using mean progress functions were developed for increasing the performance of radionuclide generators. As a result of this optimisation, the separation of the daughter nuclide from its parent one should be performed at a defined optimal time to avoid the deterioration in specific activity of the daughter nuclide and wasting stand-by time of the generator, while the daughter nuclide yield is maintained to a reasonably high extent. A new characteristic parameter of the formation-decay kinetics of parent/daughter nuclide system was found and effectively used in the practice of the generator production and utilisation. A method of "early elution schedule" was also developed for increasing the daughter nuclide production yield and specific radioactivity, thus saving the cost of the generator and improving the quality of the daughter radionuclide solution. These newly developed optimisation methods in combination with an integrated elution-purification-concentration system of radionuclide generators recently developed is the most suitable way to operate the generator effectively on the basis of economic use and improvement of purposely suitable quality and specific activity of the produced daughter radionuclides. All these features benefit the economic use of the generator, the improved quality of labelling/scan, and the lowered cost of nuclear medicine procedure. Besides, a new method of quality control protocol set-up for post-delivery test of radionuclidic purity has been developed based on the relationship between gamma ray spectrometric detection limit, required limit of impure radionuclide activity and its measurement certainty with respect to optimising decay/measurement time and product sample activity used for QC quality control. The optimisation ensures a certainty of measurement of the specific impure radionuclide and avoids wasting the useful amount of valuable purified/concentrated daughter nuclide product. This process is important for the spectrometric measurement of very low activity of impure radionuclide contamination in the radioisotope products of much higher activity used in medical imaging and targeted radiotherapy.

Influence of metal ions on the 68Ga-labeling of DOTATATE.

The influence of metal cations (Fe³âº, Fe²âº, In³âº, Cu²âº, Ca²âº, Al³âº, Co²âº, Lu³âº, Ni²âº, Pb²âº, Ti4⁺, Y³âº, Yb³âº, Zn²âº, and Zr4⁺) on the radiolabeling yield of [68Ga(DOTATATE)] was evaluated. Our most important observation was that, within our experimental limit, the metal ion/ligand ratio plays a critical role on the influence of most metal ions. More in-depth studies, with Cu²âº and Fe³âº, revealed that reaction temperature and concentration changes have little effect, but speciation changes with pH are crucial. Furthermore, we found that [68Ga(DOTATATE)] is stable in the presence of high concentrations of Fe³âº, Zn²âº and Pb²âº, but transmetalates with Cu²âº at 95°C.

(68)Ga generator integrated system: elution-purification-concentration integration.

A (68)Ge/(68)Ga generator combined with an automated (68)Ga eluate purification-concentration unit [radioisotope generator integrated system (RADIGIS)], specially designed for (68)Ga processing (RADIGIS-(68)Ga), was developed. The high-stability sorbents of a nanocrystalline structure Zr-Ti ceramic matrix were used for immobilizing the (68)Ge, and the (68)Ga was eluted from the sorbent column with 3.5 mL 0.05-0.1 M HCl solution following an optimized (68)Ga-elution schedule. The (68)Ge breakthrough <10(-3)% and no (68)Ge zone spreading/drift found in PET imaging of the (68)Ga generator column prove the excellent performance of the sorbents. (68)Ga eluate was purified on a small column of salt-form ion exchange resin using an aqueous alcohol solution mixture of hydrochloric and ascorbic acids, and halide salts. An alkali solution was used for stripping (68)Ga from the ion exchange resin column to obtain a purified (68)Ga solution, which is conditioned with acidic solution to obtain a final (68)Ga product in either 0.75 mL 0.5 M NaCl solution of pH 3-4 or 0.5 M sodium acetate or citrate solution of pH 5. The (68)Ge content in purified (68)Ga solution was <10(-6)%. An insignificant metallic contamination including (68)Zn found in the (68)Ga solution and its alkalinity-acidity were evaluated with respect to (68)Ga radiolabeling efficacy for DOTATATE and DOTATOC ligands. Quality control protocols were also developed to evaluate the quality of (68)Ga solution.

Specific radioactivity of neutron induced radioisotopes: assessment methods and application for medically useful 177Lu production as a case.

The conventional reaction yield evaluation for radioisotope production is not sufficient to set up the optimal conditions for producing radionuclide products of the desired radiochemical quality. Alternatively, the specific radioactivity (SA) assessment, dealing with the relationship between the affecting factors and the inherent properties of the target and impurities, offers a way to optimally perform the irradiation for production of the best quality radioisotopes for various applications, especially for targeting radiopharmaceutical preparation. Neutron-capture characteristics, target impurity, side nuclear reactions, target burn-up and post-irradiation processing/cooling time are the main parameters affecting the SA of the radioisotope product. These parameters have been incorporated into the format of mathematical equations for the reaction yield and SA assessment. As a method demonstration, the SA assessment of 177Lu produced based on two different reactions, 176Lu (n,γ)177Lu and 176Yb (n,γ) 177Yb (ß- decay) 177Lu, were performed. The irradiation time required for achieving a maximum yield and maximum SA value was evaluated for production based on the 176Lu (n,γ)177Lu reaction. The effect of several factors (such as elemental Lu and isotopic impurities) on the 177Lu SA degradation was evaluated for production based on the 176Yb (n,γ) 177Yb (ß- decay) 177Lu reaction. The method of SA assessment of a mixture of several radioactive sources was developed for the radioisotope produced in a reactor from different targets.

Alternative method for (64)Cu radioisotope production.

The method for (64)Cu production based on a (64)Ni target using an 18MeV proton energy beam was developed. The studies on the optimisation of targetry for the 18MeV proton bombardments were performed in terms of the cost-effective target utilisation and purity of the (64)Cu product. The thickness-specific (64)Cu yield (microCi/(microA x microm)) was introduced into the optimisation calculation with respect to cost-effective target utilisation. A maximum target utilisation efficacy factor (TUE) was found for the proton energy range of 2.5-13MeV with corresponding target thickness of 36.2microm. With the optimised target thickness and proton energy range, the (64)Ni target thickness saving of 45.6% was achieved, while the overall (64)Cu yield loss is only 23.9%, compared to the use of the whole effective proton energy range of 0-18MeV with target thickness of 66.6microm. This optimisation has the advantage of reducing the target amount to a reasonable level, and therefore the cost of the expensive (64)Ni target material. The (64)Ni target electroplated on the Au-Tl multi layer coated Cu-substrate was a new and competent design for an economic production of high quality (64)Cu radioisotope using an 18MeV proton energy cyclotron or a 30MeV cyclotron with proton beam adjustable to 18MeV. In this design, the Au coating layer plays a role of protection of "cold" Cu leakage from the Cu substrate and Tl serves to depress the proton beam energy (from 18MeV to the energy optimised value 13MeV). The ion exchange chromatographic technique with a gradient elution was applied to improve the (64)Cu separation with respect to reducing the processing time and control of (64)Cu product quality.

Development of a titanium tungstate-based 188W/188Re gel generator using tungsten of natural isotopic abundance.

The feasibility of developing titanium tungstate-based 188W/188Re gel generator using tungsten of natural isotopic abundance irradiated in a moderate flux reactor has been investigated. Influence of temperature, pH and eluent concentration on generator performance was studied. It was found that "post-formed" approach allows to construct gel generators with elution performance and 188Re elution yields very close to those of conventional alumina 188W/188Re generator. Curie-level 185W radionuclidic impurity presents a challenge during the processing of target material and subsequent elution of the generator. In the future use of semi-enriched with 186W target material (50-60% enrichment) would be beneficial in the development of titanium tungstate-based 188W/188Re gel generators.