Introduction 68Ge (t½ = 270.8 d, 100% EC) is an important radionuclide for two reasons: 1) once in equilib-rium with its daughter nuclide 68Ga (t½ = 68 min, 89 % β+, 3 % 1077 keV γ), it can be used as a positron source for attenuation correction and calibration of PET/MRI scanners; and 2) it can be employed as a generator of 68Ga for radiophar-maceutical preparation. Most isotope production facilities produce it using natural gallium (60.1% 69Ga, 39.9% 71Ga, melting point: 39 °C) as target material for proton bombardment at energies > 11.5 MeV, the threshold energy for 69Ga(p,2n)68Ge [1]. A maximum cross section of ~330 mb for natGa(p,x)68Ge occurs at ~20 MeV [1], hence proton energies in this neighborhood are mandatory for large scale production. Galli-um targetry is challenging due to its low melting point and corrosivity, hence compounds such as Ga2O3 (melting point: 1900 °C) or GaxNiy alloys (melting points > 800 °C) [2], have been used as target compounds [3,4,5]. The separation chem-istry technique employed by large-scale produc-tion facilities is liquid-liquid extraction using CCl4 [6,7]. In this work, two simple methods for GaxNiy alloy preparation are presented as well as a simple germanium separation procedure using a commercially available extraction resin. Material and Methods GaxNiy alloys were prepared by two methods (A,B). A) electrodeposition over 1.3 cm2 of a gold disk substrate. Ga2O3 and NiSO4.6H2O were dis-solved in a mixture of (27%) H2SO4 and NH4OH at pH 1.5 in a 3:2 mass ratio so that the Ga:Ni molar ratio was 4:1. The solution was then transferred to a 15 mL plating cell, in which a current of 29 mA/cm2 was applied with a platinum anode at 1 cm from the gold surface. B) Ga pellets were fused together with Ni powder at different Ga:Ni molar ratios using an induction furnace (EIA Power Cube 45/900). The resulting alloy pellets were then rolled to foils using a jeweler’s mill pressed between Nb foils to avoid contamination. Target irradiations were performed on a GE PETtrace at 16 MeV protons. The electroplated alloys were mounted on a custom-made solid target irradiation system with direct water-jet cooling applied to the backside of the gold disk. The alloy foils were placed on top of in a 1.2 cm diameter, 406 μm deep pocket made of Nb and sealed against a 51 μm Nb foil using a teflon O-ring. The alloys were in direct contact with the Nb foil to allow thermal conduction. At the rear of the Nb pocket is a water-cooling stream to transfer heat convectively during irradiation. Ge separation was achieved based on the difference in distribution coefficients between Ge, Ga, Zn, Cu, Ni and Co at different HNO3 molarities in DGA resin (Triskem International). Initial tests on the resin were performed after two pilot irradiations on natural gallium (a,b). a) 16 MeV protons were directed downward on an external beam-line (−30 °) onto 640 mg of molten elemental natGa pooled on a water-cooled niobium support. b) 330 mg natGa pellet was melted in the same Nb pocket well used with the alloys and was also sealed against a 51 μm Nb foil. The irradiated gallium was left to decay for 2 weeks and then was dissolved in 6 mL of concentrated HNO3. The solution was then passed through 200 mg of DGA resin packed in a 5 mm diameter column at a flow rate of 1.1 mL/min. A separation profile for Ge, Ga and Zn was obtained by collecting 0.2–1.0 mL fractions, which were analyzed by gamma ray spectroscopy on a HPGe detector. Two thick NiGa4 foils have been irradiated, one for 69Ge production and for radiocobalt, from 58Ni(p,α), separation quantification; and the other one for 68Ge production with the idea of preparing a mini-generator (< 13 MBq) of 68Ga for local use in phantom imaging work and animal studies. Results and Conclusion A) Each electroplating batch consisted of 66.5 ± 2.9 mg of Ga2O3 mixed with 44.9 ± 3.6 mg of NiSO4.6H2O (n = 9) in the 15 mL plating cell. Higher concentrations resulted in inefficient electroplating yields due to precipitation. 66 ± 6 % of the total Ga+Ni mass in solution, that is 39.5 ± 3.3 mg of Ga-Ni was deposited after 3 d. Three plating batches over one disk resulted in a maximum target thickness of 86.7 mg/cm2. A fourth batch did not add any significant amount of alloy and salt precipitation became a problem. The electroplated surface looked homogeneous at 10× magnification on a microscope and the targets were able to withstand up to 30 μA without presenting any dark spots. B) Alloys with Ga:Ni molar ratios of 1.0, 2.0, 2.9, 3.7 and 5.2 were fused by induction heating. TABLE 1 summarizes the results from manipulating these foils. These alloys were analyzed by X-ray fluorescence using a 109Cd excitation source quantifying the x-rays peaks: 9.26 keV for Ga and 7.48 keV for Ni. A linear relationship between the ratio of count rates of these two peaks to the alloy Ga:Ni molar ratio was found and employed for the characterization of the electroplated Ga-Ni layers. Results from the irradiations over natGa on Nb supports are presented in TABLE 2. TABLE 3 presents the results from irradiating two thick NiGa4 foils made by induction heating. Figure 1 contains the separation profile with DGA. 91% of the 68Ge is eluted in 2 mL of de-ionized water. We developed two simple methods for NiGa4 alloy manufacture. With a melting point > 800 °C and 80% presence of natGa, it is a more convenient target for 68Ge production compared to Ga encapsulated in Nb. The separation method based on the extraction resin DGA yields similar results as the liquid-liquid extraction method mentioned in [6,7], but we believe this is a more convenient method since it only requires a single trap-and-release step and not many extraction steps.

Há mais de 50 anos os geradores de 68Ge/68Ga vêm sendo desenvolvidos, obtendo o 68Ga sem a necessidade da instalação de um cíclotron próximo à radiofarmácia ou ao centro hospitalar que tenha um PET/CT. O 68Ga é um emissor de pósitron com baixa emissão de fóton (&beta;+, 89%, 1077 keV) e meia vida de 67,7 minutos, compatível com a farmacocinética de moléculas de baixo peso molecular, como peptídeos e fragmentos de anticorpos. Além disso, a química do Ga permite a ligação estável com agentes quelantes acoplados com peptídeos, como o DOTA. Todas estas características do 68Ga aliado a tecnologia PET/CT permitiram avanços em imagem molecular, como no diagnóstico de doenças de origem neuroendócrina. Entretanto, o eluato de 68Ga proveniente dos geradores de 68Ge/68Ga comerciais, ainda contém altos níveis de contaminantes, como o 68Ge e outros metais que competem quimicamente com o 68Ga, como o Fe3+ e Zn2+ e, como consequência, há redução do rendimento de marcação com biomoléculas. Quanto menor a quantidade de impurezas no eluato, a competição entre o peptídeo radiomarcado e peptídeo não marcado será menor e a qualidade de imagem será melhor, por isso existe a necessidade de diminuir a quantidade destes metais. Portanto, os objetivos deste trabalho são avaliar os métodos de purificação do 68Ga para a marcação de biomoléculas, com ênfase no estudo das impurezas químicas presentes nos radioisótopos primários, e desenvolver um método de purificação inédito. Diversos métodos de purificação foram estudados. Na purificação em resina catiônica tradicional e comercial, em que o 68Ga é adsorvido em resina catiônica e eluído em uma solução de acetona/ácido, a resina utilizada não é disponível comercialmente. Várias resinas catiônicas foram testadas simulando o processo comercial, e o uso das menores partículas da resina catiônica AG50W-X4 (200-400 mesh) foi a que apresentou os melhores resultados. Um método inovador foi a cromatografia por extração, onde o éter diisopropílico é adsorvido em resina XAD 16 e o 68Ga eluído em água deionizada. Apesar dos resultados de recuperação do 68Ga e a separação entre o 68Ga e o 65Zn terem sido bons, não houve reprodutibilidade na purificação dos metais. O método mais promissor e inédito foi a purificação do 68Ga em resina catiônica em meio básico que apresentou bons resultados, principalmente em relação à redução do Zn (98 ± 2)%, o contaminante químico encontrado em maior abundância no eluato de 68Ga. A redução total de impurezas foi (95 ± 4)%. Os peptídeos DOTATOC/DOTATATO foram marcados com o 68Ga em três diferentes formas: purificado em meio básico, por extração por solventes e sem a purificação prévia, o melhor resultado de rendimento de marcação do 68Ga DOTATATO foi obtido após a purificação do 68Ga em meio básico, comprovando a eficiência do processo. / For more than fifty years, the long-lived 68Ge/68Ga generators have been in development, obtaining 68Ga without the need of having in house cyclotron, which is a considerable convenience for PET centers that have no nearby cyclotrons. 68Ga decays 89% by positron emission and low photon emission (1077 keV) and the physical half life of 67.7 minutes is compatible with the pharmacokinetics of low biomolecular weight substances like peptides and antibody fragments. Moreover, its established metallic chemistry allows it to be stably bound to the carrier peptide sequence via a suitable bifunctional chelator, such as DOTA. All these reasons together with the technology of PET/CT allowed advances in molecular imaging, in particular in the diagnosis of neuroendocrine diseases. However, the eluate from the commercial 68Ge/68Ga generators still contains high levels of long lived 68Ge, besides other metallic impurities, which competes with 68Ga with a consequent reduction of the labeling yield of biomolecules, such as Fe3+ and Zn2+. Thus, the lower the amount of impurities in the eluate, the competition between the radiolabeled and unlabeled peptide by the receptor will be smaller and the quality of imaging will be better, a subsequent purification step is needed after the generator elution. The aim of this work is to evaluate different purifications methods of 68Ga to label biomolecules, with emphasis on the study of the chemical impurities contained in the eluate and to develop a new purification method. Several purification methods were studied. Many cationic resin were tested simulating the commercial process. 68Ga is adsorbed in cationic resin, which is not commercial available and eluted in acid/acetone solution. The use of minor particles of cationic resin AG50W-X4 (200-400 mesh) showed the best results. An innovate method was the extraction chromatography, wich is based on the absorption of diisopropyl ether in XAD 16 and 68Ga recovery in deionized water. Although the results regarding to 68Ga recovery and the radiochemical separation between 68Ga and 65Zn were excellent, there was no reproducibility on the purification of metals. The most promising and innovative method was the 68Ga purification performed by cationic resin in basic media, which presented the best results, especially regarding the Zn reduction (98 ± 2)%, the chemical contaminant found in great abundance in 68Ga eluate. The total impurities reduction was (95 ± 4)%. The peptides DOTATOC/DOTATATE were labeled 68Ga in three different forms: purified 68Ga in basic solution, through solvent extraction and no purified 68Ga. The best result was achieved with DOTATATE labeling with purified 68Ga in basic media, proving the purification process efficiency.

Há mais de 50 anos os geradores de 68Ge/68Ga vêm sendo desenvolvidos, obtendo o 68Ga sem a necessidade da instalação de um cíclotron próximo à radiofarmácia ou ao centro hospitalar que tenha um PET/CT. O 68Ga é um emissor de pósitron com baixa emissão de fóton (&beta;+, 89%, 1077 keV) e meia vida de 67,7 minutos, compatível com a farmacocinética de moléculas de baixo peso molecular, como peptídeos e fragmentos de anticorpos. Além disso, a química do Ga permite a ligação estável com agentes quelantes acoplados com peptídeos, como o DOTA. Todas estas características do 68Ga aliado a tecnologia PET/CT permitiram avanços em imagem molecular, como no diagnóstico de doenças de origem neuroendócrina. Entretanto, o eluato de 68Ga proveniente dos geradores de 68Ge/68Ga comerciais, ainda contém altos níveis de contaminantes, como o 68Ge e outros metais que competem quimicamente com o 68Ga, como o Fe3+ e Zn2+ e, como consequência, há redução do rendimento de marcação com biomoléculas. Quanto menor a quantidade de impurezas no eluato, a competição entre o peptídeo radiomarcado e peptídeo não marcado será menor e a qualidade de imagem será melhor, por isso existe a necessidade de diminuir a quantidade destes metais. Portanto, os objetivos deste trabalho são avaliar os métodos de purificação do 68Ga para a marcação de biomoléculas, com ênfase no estudo das impurezas químicas presentes nos radioisótopos primários, e desenvolver um método de purificação inédito. Diversos métodos de purificação foram estudados. Na purificação em resina catiônica tradicional e comercial, em que o 68Ga é adsorvido em resina catiônica e eluído em uma solução de acetona/ácido, a resina utilizada não é disponível comercialmente. Várias resinas catiônicas foram testadas simulando o processo comercial, e o uso das menores partículas da resina catiônica AG50W-X4 (200-400 mesh) foi a que apresentou os melhores resultados. Um método inovador foi a cromatografia por extração, onde o éter diisopropílico é adsorvido em resina XAD 16 e o 68Ga eluído em água deionizada. Apesar dos resultados de recuperação do 68Ga e a separação entre o 68Ga e o 65Zn terem sido bons, não houve reprodutibilidade na purificação dos metais. O método mais promissor e inédito foi a purificação do 68Ga em resina catiônica em meio básico que apresentou bons resultados, principalmente em relação à redução do Zn (98 ± 2)%, o contaminante químico encontrado em maior abundância no eluato de 68Ga. A redução total de impurezas foi (95 ± 4)%. Os peptídeos DOTATOC/DOTATATO foram marcados com o 68Ga em três diferentes formas: purificado em meio básico, por extração por solventes e sem a purificação prévia, o melhor resultado de rendimento de marcação do 68Ga DOTATATO foi obtido após a purificação do 68Ga em meio básico, comprovando a eficiência do processo. / For more than fifty years, the long-lived 68Ge/68Ga generators have been in development, obtaining 68Ga without the need of having in house cyclotron, which is a considerable convenience for PET centers that have no nearby cyclotrons. 68Ga decays 89% by positron emission and low photon emission (1077 keV) and the physical half life of 67.7 minutes is compatible with the pharmacokinetics of low biomolecular weight substances like peptides and antibody fragments. Moreover, its established metallic chemistry allows it to be stably bound to the carrier peptide sequence via a suitable bifunctional chelator, such as DOTA. All these reasons together with the technology of PET/CT allowed advances in molecular imaging, in particular in the diagnosis of neuroendocrine diseases. However, the eluate from the commercial 68Ge/68Ga generators still contains high levels of long lived 68Ge, besides other metallic impurities, which competes with 68Ga with a consequent reduction of the labeling yield of biomolecules, such as Fe3+ and Zn2+. Thus, the lower the amount of impurities in the eluate, the competition between the radiolabeled and unlabeled peptide by the receptor will be smaller and the quality of imaging will be better, a subsequent purification step is needed after the generator elution. The aim of this work is to evaluate different purifications methods of 68Ga to label biomolecules, with emphasis on the study of the chemical impurities contained in the eluate and to develop a new purification method. Several purification methods were studied. Many cationic resin were tested simulating the commercial process. 68Ga is adsorbed in cationic resin, which is not commercial available and eluted in acid/acetone solution. The use of minor particles of cationic resin AG50W-X4 (200-400 mesh) showed the best results. An innovate method was the extraction chromatography, wich is based on the absorption of diisopropyl ether in XAD 16 and 68Ga recovery in deionized water. Although the results regarding to 68Ga recovery and the radiochemical separation between 68Ga and 65Zn were excellent, there was no reproducibility on the purification of metals. The most promising and innovative method was the 68Ga purification performed by cationic resin in basic media, which presented the best results, especially regarding the Zn reduction (98 ± 2)%, the chemical contaminant found in great abundance in 68Ga eluate. The total impurities reduction was (95 ± 4)%. The peptides DOTATOC/DOTATATE were labeled 68Ga in three different forms: purified 68Ga in basic solution, through solvent extraction and no purified 68Ga. The best result was achieved with DOTATATE labeling with purified 68Ga in basic media, proving the purification process efficiency.

Introduction A method based on neutron activation is being developed to assist in resolving discrepancies between the expected yield and actual yield of radionuclides produced with the vertical-beam target station (VBTS) at iThemba LABS. The VBTS is routinely employed for multi-Ci batch productions of the radionuclide pairs 22Na/68Ga and 82Sr/68Ga using standardized natMg/natGa and natRb/natGa tandem targets, respectively [1]. The metal-clad target discs are bombarded with a primary beam of 66 MeV protons at an intensity of nominally 250 µA. The encapsulation materials are either Nb (for Mg and Ga) or stainless steel (for Rb) which serve to contain the molten target materials during bombardment and act as a barrier to the high-velocity cooling water which surrounds the targets in a 4π geometry. The natRb/natGa targets are typically bombarded according to a two-week cycle while natMg/natGa targets are bombarded on an ad-hoc basis, depending on a somewhat unpredictable 22Na demand. A too-large deviation between expected yield and actual yield has at times plagued this programme. These deviations can manifest both as an apparent loss or an apparent gain (relative to the expected yield) by up to about 15% in either direction. The resulting uncertainty of up to 30% (in the worst case) from one production batch to another can be costly and is unacceptable in a large-scale production regimen. This phenomenon is believed to be brought about by two types of problems: (1) Production losses, e.g. during the radio-chemical separation process or incomplete recovery of activated target material during the decapsulation step. (2) Incorrect values obtained for the accumulated proton charge. A problem of type (1) will always result in a loss of yield. A problem of type (2) can manifest as an apparent loss or gain. In an effort to get a handle on this second type of problem, neutron activation of suitable material samples, embedded in a target holder, is being investigated as an independent indicator of the total yield. For this purpose, samples of Co, Mn, Ni and Zn were activated during production runs and Co was found to be the most appropriate. Preliminary results will be presented after first discussing why the determination of the accumulated pro-ton charge is a problem with the VBTS. Materials and Methods The VBTS consists of a central region in which a target holder is located during bombardment as well as two half-cylindrical radiation shields which completely surround the target. The shields can be moved away from the central region on dedicated rails, e.g. when repairs or maintenance is required. FIGURE 1 shows the VBTS with the shields moved to the “open” position. As some components of the station are located below the vault floor, with the target position near floor level, it proved difficult to electrically isolate the VBTS as was done for the two horizontal-beam target stations at iThemba LABS [1]. The VBTS does not act as a Faraday cup like the other target stations. Instead, the beam current and accumulated charge is measured by means of a calibrated capacitive probe [1,2]. There appears to be a variation in the response of the capacitive probe, sensitive to the beam microstructure, in particular a dependence on the beam packet length. This problem is not yet fully resolved. FIGURE 2 (a) shows the beamstop of a VBTS target holder with several Co samples mounted on the outside as well as one each of Ni, Mn and Zn. The samples are small “tablets” with a 10 mm diameter and 1 mm thickness. The reactions of interest are 59Co(n,γ)60Co, 59Co(n,3n)57Co, nat-Ni(n,X)60Co, natNi(n,X)57Co, natZn(n,X)65Zn and 55Mn(n,2n)54Mn. The relevant half-lives are 60Co(5.271 a), 57Co(271.8 d), 65Zn(244.3 d) and 54Mn(312.2 d). The half-life should be long compared to the two-week cycle in order to reduce the dependence on the exact beam history, which is very fragmented over any production period. In this respect, 60Co is considered to be particularly attractive as its long half-life of more than 5 years leads to a negligible effect by the beam history. Note that the tandem targets, shown in FIGURE 2 (b), are mounted just upstream of the beamstop – in fact, the targets and beamstop form a single unit before being fitted into the target holder. At the end of bombardment, all samples were assayed for their characteristic γ-emissions using standard off-line γ-ray spectrometry with an HPGe detector connected to a Genie 2000 MCA. Calculations of the neutron fluence density in the central sample volume on the beamstop were also performed using the Monte Carlo radiation transport code MCNPX. For these calculations, the entire VBTS, a Rb/Ga target and the vault walls were included in the model. Results and Conclusion All samples activated significantly – copious amounts of 60Co were detected in the Co discs after a two-week run. The neutron fluence density for the case of a 250 µA, 66 MeV proton beam on a natRb/natGa tandem target is shown in FIGURE 3. The dominance of low-energy neutrons is evident, which is in part due to the large amount of paraffin-wax shielding material in close proximity to the target. While reactions such as the (n,2n) and (n,3n) would be sensitive to the more energetic part of the neutron spectrum, the (n,γ) capture reaction benefits from the large low-energy component. This explains the copious amounts of 60Co formed. It was therefore decided to only retain the central Co sample for subsequent bombardments, as shown in FIGURE 4. The first results are shown in TABLE 1. The accumulated charge as obtained from the capacitive probe (Q), the specific 60Co activity (A) at the end of bombardment (EOB), and their ratio (A/Q) are presented in the table, together with the deviation of individual ratios relative to their average for the case of the Mg/Ga tandem tar-gets only. Note that all samples were counted until the statistical uncertainties were negligible. Any systematic uncertainties are ignored at this stage as they are considered to remain the same from one batch production to another. For the sake of argument, the average value of the ratio is taken as the expected value. A positive deviation of the A/Q value is then indicative of a too-small value of the accumulated charge obtained from the capacitive probe, leading to a corresponding overproduction. Likewise, a negative value is indicative of a too-large value of the accumulated charge, leading to a corresponding underproduction. It is certainly true that the data in TABLE 1 are currently very limited. It is envisaged, however, that with time the growing database of values will assist in reducing the uncertainty in determining the accumulated charge and reduce the discrepancies between predicted and actual yields significantly. TABLE 1 illuminates the underlying problem satisfactorily. The four Mg/Ga tandem target bombardments, on identical targetry, were performed successively. The neutron activation correlates well the with actual yields, pointing directly to the current integration as the main source of error. The method already proves to be useful. An indication of an over or underprediction can be obtained prior to the target processing by recovering and measuring the Co disc. This in-formation can be used to make a decision concerning the present batch production and/or the subsequent one. One can either add beam to the present production target and/or in-crease/reduce the total beam on the subsequent production target to compensate for an expected overproduction or shortfall. In conclusion, we would like to stress that the capacitive probes show great promise and that better understanding and/or possibly some development of their signal processing algorithm may improve their ability to measure the accumulated charge to the desired accuracy. Segmented capacitive probes used at iThemba LABS and elsewhere for beam position measurement [1,3] are not affected by beam microstructure as only the ratios of the signal strengths on the different sectors are important. In this case, changes in response affect all sec-tors equally and the ratios are unaffected.

Introduction 68Ge (t½ = 270.8 d, 100% EC) is an important radionuclide for two reasons: 1) once in equilib-rium with its daughter nuclide 68Ga (t½ = 68 min, 89 % β+, 3 % 1077 keV γ), it can be used as a positron source for attenuation correction and calibration of PET/MRI scanners; and 2) it can be employed as a generator of 68Ga for radiophar-maceutical preparation. Most isotope production facilities produce it using natural gallium (60.1% 69Ga, 39.9% 71Ga, melting point: 39 °C) as target material for proton bombardment at energies > 11.5 MeV, the threshold energy for 69Ga(p,2n)68Ge [1]. A maximum cross section of ~330 mb for natGa(p,x)68Ge occurs at ~20 MeV [1], hence proton energies in this neighborhood are mandatory for large scale production. Galli-um targetry is challenging due to its low melting point and corrosivity, hence compounds such as Ga2O3 (melting point: 1900 °C) or GaxNiy alloys (melting points > 800 °C) [2], have been used as target compounds [3,4,5]. The separation chem-istry technique employed by large-scale produc-tion facilities is liquid-liquid extraction using CCl4 [6,7]. In this work, two simple methods for GaxNiy alloy preparation are presented as well as a simple germanium separation procedure using a commercially available extraction resin. Material and Methods GaxNiy alloys were prepared by two methods (A,B). A) electrodeposition over 1.3 cm2 of a gold disk substrate. Ga2O3 and NiSO4.6H2O were dis-solved in a mixture of (27%) H2SO4 and NH4OH at pH 1.5 in a 3:2 mass ratio so that the Ga:Ni molar ratio was 4:1. The solution was then transferred to a 15 mL plating cell, in which a current of 29 mA/cm2 was applied with a platinum anode at 1 cm from the gold surface. B) Ga pellets were fused together with Ni powder at different Ga:Ni molar ratios using an induction furnace (EIA Power Cube 45/900). The resulting alloy pellets were then rolled to foils using a jeweler’s mill pressed between Nb foils to avoid contamination. Target irradiations were performed on a GE PETtrace at 16 MeV protons. The electroplated alloys were mounted on a custom-made solid target irradiation system with direct water-jet cooling applied to the backside of the gold disk. The alloy foils were placed on top of in a 1.2 cm diameter, 406 μm deep pocket made of Nb and sealed against a 51 μm Nb foil using a teflon O-ring. The alloys were in direct contact with the Nb foil to allow thermal conduction. At the rear of the Nb pocket is a water-cooling stream to transfer heat convectively during irradiation. Ge separation was achieved based on the difference in distribution coefficients between Ge, Ga, Zn, Cu, Ni and Co at different HNO3 molarities in DGA resin (Triskem International). Initial tests on the resin were performed after two pilot irradiations on natural gallium (a,b). a) 16 MeV protons were directed downward on an external beam-line (−30 °) onto 640 mg of molten elemental natGa pooled on a water-cooled niobium support. b) 330 mg natGa pellet was melted in the same Nb pocket well used with the alloys and was also sealed against a 51 μm Nb foil. The irradiated gallium was left to decay for 2 weeks and then was dissolved in 6 mL of concentrated HNO3. The solution was then passed through 200 mg of DGA resin packed in a 5 mm diameter column at a flow rate of 1.1 mL/min. A separation profile for Ge, Ga and Zn was obtained by collecting 0.2–1.0 mL fractions, which were analyzed by gamma ray spectroscopy on a HPGe detector. Two thick NiGa4 foils have been irradiated, one for 69Ge production and for radiocobalt, from 58Ni(p,α), separation quantification; and the other one for 68Ge production with the idea of preparing a mini-generator (< 13 MBq) of 68Ga for local use in phantom imaging work and animal studies. Results and Conclusion A) Each electroplating batch consisted of 66.5 ± 2.9 mg of Ga2O3 mixed with 44.9 ± 3.6 mg of NiSO4.6H2O (n = 9) in the 15 mL plating cell. Higher concentrations resulted in inefficient electroplating yields due to precipitation. 66 ± 6 % of the total Ga+Ni mass in solution, that is 39.5 ± 3.3 mg of Ga-Ni was deposited after 3 d. Three plating batches over one disk resulted in a maximum target thickness of 86.7 mg/cm2. A fourth batch did not add any significant amount of alloy and salt precipitation became a problem. The electroplated surface looked homogeneous at 10× magnification on a microscope and the targets were able to withstand up to 30 μA without presenting any dark spots. B) Alloys with Ga:Ni molar ratios of 1.0, 2.0, 2.9, 3.7 and 5.2 were fused by induction heating. TABLE 1 summarizes the results from manipulating these foils. These alloys were analyzed by X-ray fluorescence using a 109Cd excitation source quantifying the x-rays peaks: 9.26 keV for Ga and 7.48 keV for Ni. A linear relationship between the ratio of count rates of these two peaks to the alloy Ga:Ni molar ratio was found and employed for the characterization of the electroplated Ga-Ni layers. Results from the irradiations over natGa on Nb supports are presented in TABLE 2. TABLE 3 presents the results from irradiating two thick NiGa4 foils made by induction heating. Figure 1 contains the separation profile with DGA. 91% of the 68Ge is eluted in 2 mL of de-ionized water. We developed two simple methods for NiGa4 alloy manufacture. With a melting point > 800 °C and 80% presence of natGa, it is a more convenient target for 68Ge production compared to Ga encapsulated in Nb. The separation method based on the extraction resin DGA yields similar results as the liquid-liquid extraction method mentioned in [6,7], but we believe this is a more convenient method since it only requires a single trap-and-release step and not many extraction steps.

Introduction A method based on neutron activation is being developed to assist in resolving discrepancies between the expected yield and actual yield of radionuclides produced with the vertical-beam target station (VBTS) at iThemba LABS. The VBTS is routinely employed for multi-Ci batch productions of the radionuclide pairs 22Na/68Ga and 82Sr/68Ga using standardized natMg/natGa and natRb/natGa tandem targets, respectively [1]. The metal-clad target discs are bombarded with a primary beam of 66 MeV protons at an intensity of nominally 250 µA. The encapsulation materials are either Nb (for Mg and Ga) or stainless steel (for Rb) which serve to contain the molten target materials during bombardment and act as a barrier to the high-velocity cooling water which surrounds the targets in a 4π geometry. The natRb/natGa targets are typically bombarded according to a two-week cycle while natMg/natGa targets are bombarded on an ad-hoc basis, depending on a somewhat unpredictable 22Na demand. A too-large deviation between expected yield and actual yield has at times plagued this programme. These deviations can manifest both as an apparent loss or an apparent gain (relative to the expected yield) by up to about 15% in either direction. The resulting uncertainty of up to 30% (in the worst case) from one production batch to another can be costly and is unacceptable in a large-scale production regimen. This phenomenon is believed to be brought about by two types of problems: (1) Production losses, e.g. during the radio-chemical separation process or incomplete recovery of activated target material during the decapsulation step. (2) Incorrect values obtained for the accumulated proton charge. A problem of type (1) will always result in a loss of yield. A problem of type (2) can manifest as an apparent loss or gain. In an effort to get a handle on this second type of problem, neutron activation of suitable material samples, embedded in a target holder, is being investigated as an independent indicator of the total yield. For this purpose, samples of Co, Mn, Ni and Zn were activated during production runs and Co was found to be the most appropriate. Preliminary results will be presented after first discussing why the determination of the accumulated pro-ton charge is a problem with the VBTS. Materials and Methods The VBTS consists of a central region in which a target holder is located during bombardment as well as two half-cylindrical radiation shields which completely surround the target. The shields can be moved away from the central region on dedicated rails, e.g. when repairs or maintenance is required. FIGURE 1 shows the VBTS with the shields moved to the “open” position. As some components of the station are located below the vault floor, with the target position near floor level, it proved difficult to electrically isolate the VBTS as was done for the two horizontal-beam target stations at iThemba LABS [1]. The VBTS does not act as a Faraday cup like the other target stations. Instead, the beam current and accumulated charge is measured by means of a calibrated capacitive probe [1,2]. There appears to be a variation in the response of the capacitive probe, sensitive to the beam microstructure, in particular a dependence on the beam packet length. This problem is not yet fully resolved. FIGURE 2 (a) shows the beamstop of a VBTS target holder with several Co samples mounted on the outside as well as one each of Ni, Mn and Zn. The samples are small “tablets” with a 10 mm diameter and 1 mm thickness. The reactions of interest are 59Co(n,γ)60Co, 59Co(n,3n)57Co, nat-Ni(n,X)60Co, natNi(n,X)57Co, natZn(n,X)65Zn and 55Mn(n,2n)54Mn. The relevant half-lives are 60Co(5.271 a), 57Co(271.8 d), 65Zn(244.3 d) and 54Mn(312.2 d). The half-life should be long compared to the two-week cycle in order to reduce the dependence on the exact beam history, which is very fragmented over any production period. In this respect, 60Co is considered to be particularly attractive as its long half-life of more than 5 years leads to a negligible effect by the beam history. Note that the tandem targets, shown in FIGURE 2 (b), are mounted just upstream of the beamstop – in fact, the targets and beamstop form a single unit before being fitted into the target holder. At the end of bombardment, all samples were assayed for their characteristic γ-emissions using standard off-line γ-ray spectrometry with an HPGe detector connected to a Genie 2000 MCA. Calculations of the neutron fluence density in the central sample volume on the beamstop were also performed using the Monte Carlo radiation transport code MCNPX. For these calculations, the entire VBTS, a Rb/Ga target and the vault walls were included in the model. Results and Conclusion All samples activated significantly – copious amounts of 60Co were detected in the Co discs after a two-week run. The neutron fluence density for the case of a 250 µA, 66 MeV proton beam on a natRb/natGa tandem target is shown in FIGURE 3. The dominance of low-energy neutrons is evident, which is in part due to the large amount of paraffin-wax shielding material in close proximity to the target. While reactions such as the (n,2n) and (n,3n) would be sensitive to the more energetic part of the neutron spectrum, the (n,γ) capture reaction benefits from the large low-energy component. This explains the copious amounts of 60Co formed. It was therefore decided to only retain the central Co sample for subsequent bombardments, as shown in FIGURE 4. The first results are shown in TABLE 1. The accumulated charge as obtained from the capacitive probe (Q), the specific 60Co activity (A) at the end of bombardment (EOB), and their ratio (A/Q) are presented in the table, together with the deviation of individual ratios relative to their average for the case of the Mg/Ga tandem tar-gets only. Note that all samples were counted until the statistical uncertainties were negligible. Any systematic uncertainties are ignored at this stage as they are considered to remain the same from one batch production to another. For the sake of argument, the average value of the ratio is taken as the expected value. A positive deviation of the A/Q value is then indicative of a too-small value of the accumulated charge obtained from the capacitive probe, leading to a corresponding overproduction. Likewise, a negative value is indicative of a too-large value of the accumulated charge, leading to a corresponding underproduction. It is certainly true that the data in TABLE 1 are currently very limited. It is envisaged, however, that with time the growing database of values will assist in reducing the uncertainty in determining the accumulated charge and reduce the discrepancies between predicted and actual yields significantly. TABLE 1 illuminates the underlying problem satisfactorily. The four Mg/Ga tandem target bombardments, on identical targetry, were performed successively. The neutron activation correlates well the with actual yields, pointing directly to the current integration as the main source of error. The method already proves to be useful. An indication of an over or underprediction can be obtained prior to the target processing by recovering and measuring the Co disc. This in-formation can be used to make a decision concerning the present batch production and/or the subsequent one. One can either add beam to the present production target and/or in-crease/reduce the total beam on the subsequent production target to compensate for an expected overproduction or shortfall. In conclusion, we would like to stress that the capacitive probes show great promise and that better understanding and/or possibly some development of their signal processing algorithm may improve their ability to measure the accumulated charge to the desired accuracy. Segmented capacitive probes used at iThemba LABS and elsewhere for beam position measurement [1,3] are not affected by beam microstructure as only the ratios of the signal strengths on the different sectors are important. In this case, changes in response affect all sec-tors equally and the ratios are unaffected.

>Magister Scientiae - MSc / 68Ge/68Ga generators rely on metal oxide, inorganic and organic sorbents in order to prepare radionuclides useful for clinical applications. The requirements for 68Ge/68Ga generators are that the 68Ga obtained from the 68Ge loaded column should be optimally suited for the routine synthesis of 68Ga-labelled radiopharmaceuticals, that the separation of the 68Ga daughter from the 68Ge parent should happen easily, with a high yield of separation, a low specific volume of 68Ga and should not contain trace elements owing to the solubility of the metal oxide sorbent. Beginning with a metal oxide preparation and continuing through recent developments, several approaches for processing generator derived 68Ga have altered the production of 68Ge/68Ga generators. Still, the effects of sorbent modification on the properties of 68Ge/68Ga radionuclide generator systems are not necessarily optimally designed for direct application in a medical context. The objective of this research was to analyze and document characteristics of Titanium Oxide (TiO2) sorbents relevant to processing of a 68Ge/68Ga generator that is able to produce 68Ga eluates that are adequate for clinical requirements. Interest was shown in TiO2 based 68Ge/68Ga generators by a number of overseas companies for tumour imaging using 68Ga-labelled 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA)-conjugated peptides. While a method involving production of the 68Ga radionuclide using TiO2 metal oxide had been published, problems with the production persisted. A method, using TiO2 metal oxide for ion exchange chromatography, was devised in this study to produce the 68Ga radionuclide, with the aim of being adopted for production purposes. The study focuses on the development of a dedicated procedure for the achievement of sufficient 68Ga yield along with low 68Ge breakthrough and low metallic impurities. Literature from 1970 to 2011 was reviewed to assess the radiochemical aspects of the 68Ga production and processing thereof. Various commercially available TiO2 metal oxides were characterized by subjecting the materials to x-ray diffraction (XRD), x-ray fluorescence (XRF) and scanning electron microscopy (SEM) for quantitative and qualitative analysis.

Introduction: Targeted pharmaceuticals for labelling with radio-isotopes for very specific imaging (and possibly later for targeted therapy) play a major role in Theranostics which is currently an important topic in Nuclear Medicine as well as personalised medicine. There was a need for a very specific lung cancer radiopharmaceutical that would specifically be uptaken in integrin 3 expression cells to image patients using a Positron Emission Tomography- Computed Tomography (PET-CT) scanner. Background and problem statement: Cold kits of c (RGDyK)–SCN-Bz-NOTA were kindly donated by Seoul National University (SNU) to help meet Steve Biko Hospital’s need for this type of imaging. These cold kits showed great results internationally in labelling with a 0.1 M 68Ge/68Ga generator (t1/2 of 68Ge and 68Ga are 270.8 days and 67.6 min, respectively). However the same cold kits failed to show reproducible radiolabeling with the 0.6 M generator manufactured under cGMP conditions at iThemba LABS, Cape Town and distributed by IDB Holland, the Netherlands. Materials and methods: There was therefore a need for producing an in-house NOTA-RGD kit that would enable production of clinical 68Ga-NOTA-RGD in high yields from the IDB Holland/iThemba LABS generator. Quality control included ITLC in citric acid to observe labelling efficiency as well as in sodium carbonate to evaluate colloid formation. HPLC was also performed at iThemba LABS as well as Necsa (South African Nuclear Energy Corporation). RGD was obtained from Futurechem, Korea. Kit mass integrity was determined by testing labelling efficiency of 10, 30 and 60 μg of RGD per cold kit. The RGD was buffered with sodium acetate trihydrate. The original kits were dried in a desiccator and in later studies only freeze dried. Manual labelling was also tested. The radiolabelled in-house kit’s ex vivo biodistribution in healthy versus tumour mice were examined by obtaining xenografts. The normal biodistribution was investigated in three vervet monkeys by doing PET-CT scans on a Siemens Biograph TP 40 slice scanner. Results: Cold kit formulation radiolabeling and purification methods were established successfully and SOPs (standard operating procedures) created. HPLC results showed highest radiochemical purity in 60 μg cold kit vials. 68Ga-NOTA-RGD showed increased uptake in tumours of tumour bearing mouse. The cold kit also showed normal distribution according to literature with fast blood clearance and excretion through kidneys into urine, therefore making it a suitable radiopharmaceutical for clinical studies. Conclusion: The in-house prepared cold kit with a 4 month shelf-life was successfully tested in mice and monkeys. / MSc (Pharmaceutics), North-West University, Potchefstroom Campus, 2014

Introduction: Targeted pharmaceuticals for labelling with radio-isotopes for very specific imaging (and possibly later for targeted therapy) play a major role in Theranostics which is currently an important topic in Nuclear Medicine as well as personalised medicine. There was a need for a very specific lung cancer radiopharmaceutical that would specifically be uptaken in integrin 3 expression cells to image patients using a Positron Emission Tomography- Computed Tomography (PET-CT) scanner. Background and problem statement: Cold kits of c (RGDyK)–SCN-Bz-NOTA were kindly donated by Seoul National University (SNU) to help meet Steve Biko Hospital’s need for this type of imaging. These cold kits showed great results internationally in labelling with a 0.1 M 68Ge/68Ga generator (t1/2 of 68Ge and 68Ga are 270.8 days and 67.6 min, respectively). However the same cold kits failed to show reproducible radiolabeling with the 0.6 M generator manufactured under cGMP conditions at iThemba LABS, Cape Town and distributed by IDB Holland, the Netherlands. Materials and methods: There was therefore a need for producing an in-house NOTA-RGD kit that would enable production of clinical 68Ga-NOTA-RGD in high yields from the IDB Holland/iThemba LABS generator. Quality control included ITLC in citric acid to observe labelling efficiency as well as in sodium carbonate to evaluate colloid formation. HPLC was also performed at iThemba LABS as well as Necsa (South African Nuclear Energy Corporation). RGD was obtained from Futurechem, Korea. Kit mass integrity was determined by testing labelling efficiency of 10, 30 and 60 μg of RGD per cold kit. The RGD was buffered with sodium acetate trihydrate. The original kits were dried in a desiccator and in later studies only freeze dried. Manual labelling was also tested. The radiolabelled in-house kit’s ex vivo biodistribution in healthy versus tumour mice were examined by obtaining xenografts. The normal biodistribution was investigated in three vervet monkeys by doing PET-CT scans on a Siemens Biograph TP 40 slice scanner. Results: Cold kit formulation radiolabeling and purification methods were established successfully and SOPs (standard operating procedures) created. HPLC results showed highest radiochemical purity in 60 μg cold kit vials. 68Ga-NOTA-RGD showed increased uptake in tumours of tumour bearing mouse. The cold kit also showed normal distribution according to literature with fast blood clearance and excretion through kidneys into urine, therefore making it a suitable radiopharmaceutical for clinical studies. Conclusion: The in-house prepared cold kit with a 4 month shelf-life was successfully tested in mice and monkeys. / MSc (Pharmaceutics), North-West University, Potchefstroom Campus, 2014

<p>The positron emitting radionuclide ⁶⁸Ga (T_1/2 = 68 min) might become of practical interest for clinical positron emission tomography (PET). The metallic cation, ⁶⁸Ga(III), is suitable for complexation with chelators, either naked or conjugated with biological macromolecules. Such labelling procedures require pure and concentrated preparations of ⁶⁸Ga(III), which cannot be sufficiently fulfilled by the presently available ⁶⁸Ge/⁶⁸Ga generator eluate. This thesis presents methods to increase the concentration and purity of ⁶⁸Ga obtained from a commercial ⁶⁸Ge/⁶⁸Ga generator. The use of the preconcentrated and purified ⁶⁸Ga eluate along with microwave heating allowed quantitative ⁶⁸Ga-labelling of peptide conjugates within 15 min. The specific radioactivity of the radiolabelled peptides was improved considerably compared to previously applied techniques using non-treated generator eluate and conventional heating. A commercial ⁶⁸Ge/⁶⁸Ga generator in combination with the method for preconcentration/purification and microwave heated labelling might result in an automated device for ⁶⁸Ga-based radiopharmaceutical kit production with quantitative incorporation of ⁶⁸Ga(III).</p><p>Macromolecules were labelled with ⁶⁸Ga(III) either directly or via a chelator. The bifunctional chelator, DOTA, was conjugated in solution to peptides, an antibody and oligonucleotides. The peptides had varied pI values, constitution, and length ranging from 8 to 53 amino acid residues. The oligonucleotides were of various sequences and length with modifications in backbone, sugar moiety and both 3' and 5' ends with a molecular weight up to 9.8 kDa. The bioconjugates were labeled with ⁶⁸Ga(III), and the resulting tracers were characterised chemically and biologically. The identity of the ⁶⁸Ga-labelled bioconjugates was verified. The tracers were found to be stable and their biological activity maintained. Specific radioactivity was shown to be an important parameter influencing the feasibility of accurate imaging data quantification.</p><p>Furthermore, ⁶⁸Ga-labelled peptide imaging was shown to be a useful tool to study peptide adsorption to microstructures in a chemical analysis device.</p>