Qucosa – Hemholtz-Zentrum Dresden-Rossendorf
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Preparation of metallic target of 100Mo for production of 99mTc in cyclotron
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Technetium-99m, the daughter of 99Mo is the most commonly used radioisotope in nuclear medicine [1–2]. Current global crisis of 99Mo supply, aging of nuclear reactors and staggering costs force the search for alternative sources of 99mTc. Radioisotope Centre POLATOM joined the IAEA Coordinated Research Project on “Accelerator-based Alternatives to Non-HEU Production of 99Mo/99mTc”. The planned outcome of this project is development of 99mTc production method using the reaction of 100Mo(p,2n)99mTc [3] in Polish cyclotron.
This work presents the results concerning preparation of 100Mo target for irradiation with protons.
Material and Methods
The manufacturing of Mo target was performed using pressing of molybdenum powder into pellets and its sintering in hydrogen atmosphere at 1600 oC [4]. For this purpose a tantalum and stainless steel plates were used as support. Several pellets using molybdenum powder with particles size of 2 µm in diameter were pressed at different values of pressure.
Results and Conclusion
The optimized parameters of pressing molyb-denum pellets with various sizes are given in TABLE 1. It was found that the pellets did not adhere neither to the tantalum nor stainless steel plates but they conducted electricity very well. Pellets prepared with higher pressure were more mechanically resistant, however application, even the highest used pressure did not ensure its satisfactory stability.
In order to improve mechanical strength, pressed Mo pellets were sintered in hydrogen atmosphere at temperature of 1600 °C. As a result of this process dimensions of Mo pellets decreased: diameter by 13 %, thickness by 12 %, weight by 1.5 %, volume by 34 % while density increased by 50 %. The changes of these parameters are associated with reduction of molybdenum oxide and removal of oxygen from intermetallic space. It was confirmed by photos of microscopic cross section of pellets before and after sintering. It was observed, that after sintering Mo pellets got a metallic form with very high hardness and mechanical strength
New gas target system for 83Rb production
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Short-lived isomer 83mKr (T½ = 1.83 h) is an ideal calibration source in several low-energy experiments like or KATRIN (determining the neutrino rest mass, monitoring high voltage stability and investigation of the main spectrometer properties) or XENON (detection of the dark matter).
The isomer 83mKr is formed by decay of 83Rb (T½ = 86.2 d) that can be produced predominantly via the reaction 84Kr(p,2n)83Rb by irradiation of natKr (57 % abundance of 84Kr).
The design and construction of the new gas target for effective production of radionuclide 83Rb as well as target processing will be shortly described.
Material and Methods
For the target design, we selected the following criteria: minimizing activation of target components; efficient cooling system allowing higher beam currents; easy handling; high life-time of the target chamber (low impact of the irradiation and radionuclide separation process on the target chamber surface and 83Rb recovery).
The target consists of three parts:
1. Water cooled aluminium (alloy EN 6082) mechanical interface for easy connection of the target to the beam line. It also serves as a beam collimator (diameter 9 mm).
2. Holder of He-cooled foils (vacuum separation foil – Havar 0.025 mm, target body window – Ti 0.1 mm).
3. Aluminium (alloy EN 6082) water cooled target body with 150mm long cone-shaped target chamber of the volume 27.1 ml. Internal surface of the chamber is nickel-coated.
The target filled with natural Kr of purity 0.9999 and absolute pressure 13 bar was irradiated on the external beam of the isochronous cyclotron U-120M of the NPI AS CR. The proton beam energy was set so that it is decreased after deg-radation in the separation foils to 25.6 MeV. Beam energy loss in the natural Kr gas filling is 9.6 MeV. The target was tested up to 25 µA beam current.
After irradiation, the target is left for a week to let the short-lived activation products to decay. Then, 83Rb is washed out from the target walls by two portions of freshly prepared de-ionized water, target is rinsed by high-purity ethanol and dried. The two portions of 83Rb aqueous solution are then connected and activity and radionuclidic purity of the product is determined via γ-spectrometry (HPGe detector). Large-distance sample-detector measurements of the target prior and after the separation are used in order to determine recovery of 83Rb.
Results and Conclusion
The new gas target for routine production of 83Rb was successfully designed, tested and im-plemented for regular 83Rb production. Six-hour irradiation with 15 µA proton beam resulted repeatedly in ca 300 MBq of 83Rb (EOB). Besides 83Rb, we identified in the separated product also 84Rb (T½ = 32.82 d) at levels ca 31 % of the 83Rb activity (EOB) and 86Rb (T½ = 18.631 d) at levels ca 8 % of the 83Rb activity (EOB). Both radionuclidic impurities do not disturb the use of 83Rb, since none of them emanates any radioactive krypton isotope. Moreover, their relative content decreases in time. Rubidium isotopes are recovered from the target almost quantitatively (98–99 %)
Development of a Krypton target for Cyclone-30 at KFSH&RC
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Krypton-81m is a radioactive gas with a half-life of 13 s, and found to be useful in many applications in nuclear medicine, particularly for lung perfusion studies and ventilations. Due to high demands for 81mKr, we have developed an automated Krypton system to be installed in one of the Cyclotron’s beamlines at King Faisal Specialist Hospital and Research Centre (KFSH&RC) and to deliver large activity of the radioactive gas.
Material and Methods
The effective cross section of producing 81Rb is between 15 and 30 MeV [1]. Therefore, range and stopping power of the effective cross section were calculated with respect to gas density of 0.0185 g/cm3. This value is equivalent to gas density at 5.0 bars at room temperature. SRIM calculations resulted in a range of 589 mm. However, due to limitation in fabricating such long target chamber, the target length is chosen to be 250 mm. Attached to the end of target body is a special water circulating flange ‘back-pool’, its purpose is to absorb the rest of the energy and protons Bragg peak. The target body is made of Aluminum with the inner part being electroplated with nickel. The target body is of conical shape. The target body is electrically isolated from other parts to allow accurate beam current reading.
Full access to the target loading/unloading steps is made through touch screen technology (FIG. 2) for user access. Additionally, the target control system is designed to be protected through chain of interlock steps. The production cycle of 81Rb is explained as follow. Target is evacuated to approximately 10−3 mbar before being filled with natKr at pressure of 5 bars. At the end of bombardment, recovery of natKr is done via cryogenic vessel. Finally, the radioactivity is washed with KCl and pushed to Hotcells through the nitrogen gas for chemistry processing. Irradiation time was approximately 30 min.
Results and Conclusion
Experimental results clearly showed a fairly good activity of 81mKr as shown in TABLE 1. In all experiments, the radionuclidic purity of 81mKr was above 99.59%. 79mKr and 79Kr were also measured with a percentage of, respectively, 0.34 and 0.07 %. Special attention has to be drawn to last experiment where the yield significantly in-creased, due to the period where the KCl left inside the target (10 min) before pushing the solution to the Hotcell
Reducing metal contamination in Cu-64 production
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In the past several years there has been a growing interest in the development of radiopharmaceuticals labeled with metallic radionuclides (Anderson et al. 1999). Of particular interest is the positron emitter Cu-64 (t½ = 12.7 h) for molecular imaging of small molecules as well as peptides and antibodies (Smith 2004). This has led us to the recent implementation of a solid target production facility using commercially available target irradiation station and chemistry modules. Routine production of Cu-64 was achieved with an average production yield of 0.32 mCi/μAh, however purification of Cu-64 has proven to be problematic; with several metallic contaminants compromising subsequent radiolabeling. We report in this work, the step by step procedure which led us to the successful production of low metal contaminant 64Cu with high specific activity and high labeling efficiency.
Material and Methods
Detailed implementation of our solid target was reported earlier (Poniger et al. 2012). A Nirta Solid Target from IBA was coupled to our 18/9 cyclotron using a 2-meter external beam line. A pneumatic solid target transfer system (STTS) designed by TEMA was use to deliver the irradiated target disks to a dedicated hotcell. Modules from IBA (Pinctada metal) were used for electroplating 64Ni onto a Ag disk and for acid dissolution and purification of the irradiated target. Typical irradiation parameters were 14.9 MeV at 35 μA for 5–6 hours with 64Ni plating’s ranging from 10–60 μm thickness at 6–12 mm. Radionuclidic purities were evaluated by gamma spectroscopy and traces of metallic impurities were determined by ICP-MS or ICP-AES. Labeling efficiency was evaluated by measuring the amount of 64Cu uptake per 20 μg of scFv-cage.
Results and Conclusion
Initial 64Cu purifications following the manufacturers recommended method resulted in high levels of Cu, Fe and Zn metal contaminants (see TABLE 1, ID 1). Note that little Ag contamination is observed nevertheless the 64Ni is plated directly on a Ag disk. After several productions, visual inspection of the module quickly revealed that the heater block used for heating the back of the Ag target disk was heavily corroded. Replacing the copper heater block with a PEEK heater block drastically reduced the levels of Cu and Fe contaminants.
Unfortunately unusually high levels of Zn were still observed regardless of the stringent conditions and ultrapure reagents used during the processing (see TABLE 1, ID 5). In our quest for answers, ICP-MS analysis of the 64Ni plating solution as well as critical stock reagents such as Milli-Q water (18 MΩ cm−1) and 30% HCl TraceSelect Ultra (Sigma) was performed (see TABLE 1, ID 2,3,4). The results were surprising, with high level of Zn found not only in the 64Ni plating solution, but as well in the HCl TraceSelect Ultra. It was hypothesized that the Pinctada’s glass bottles (Kay, 2004) used to store the reagents, especially concentrated acidic solutions were the source of Zn contamination and all glass bottles were replaced by LDPE or PFA types. Our hypothesis was confirmed by subsequent ICP-MS analysis of fresh samples of HCl TraceSelect Ultra and the 64Ni plating solution prepared/stored in plastic containers (see TABLE 1, ID 6,7). We also confirmed by ICP-MS analysis that no contamination occurred when performing a non-radioactive dissolution/purification sequence on the Pinctada module using a blank PTFE target disk in conjunction with the change to plastic reagent storage bottles (see TABLE 1, ID 8).
Initially the purification protocol was modified as described by Ometakova et al., 2012 to help reduce the co-elution of Zn contaminants with the 64Cu from the AG1-X8 resin. This change resulted in a significant amount of 64Cu eluting from the resin during the resin washing steps, so that protocol was abandoned and the protocol as described by Thieme et al., 2012 was adopted. By modifying the AG1-X8 resin washing protocol to this new method and eluting the 64Cu from with 0.1M HCl rather than Milli-Q water (see TABLE 1, ID 9), we were able to further reduce metal contaminants, especially Zn.
During the course of these experiments, the true specific activity of 64Cu increased from as low as 12 mCi/μmol of Cu (n = 2, TABLE 1, ID 1) to 649 mCi/μmol of Cu (n = 7, TABLE 1, ID 5) and finally to 4412 mCi/μmol of Cu (n = 3, TABLE 1, ID 9). In the same time, the effective specific activity increased from 0.03 ± 0.02 mCi per 20 μg of scFv-cage, to 3.7 ± 0.3 mCi per 20 g of scFv-cage with 64Cu.
In conclusion, a significant reduction in Cu, Fe and Zn contaminants was achieved when processing 64Cu using the Pinctada module: i) after replacement of the Cu heater block; ii) after elimination of glass reagent storage containers from the Pinctada module and procedures during preparation of the 64Ni plating solution and iii) after implementation of a new purification protocol (Thieme et al. 2012). Introduction of a 6M HCl wash-up cycle of the module prior to the dissolution procedure was also effective. However in recent 64Cu productions slightly elevated Ag levels have been observed and are under investigation (see TABLE 1, ID 9)
Production of Radiobromide: new Nickel Selenide target and optimized separation by dry distillation
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Radioisotopes of bromine are of special interest for nuclear medical applications. The positron emitting isotopes 75Br (T½ = 1.6 h; β+ = 75.5 %) and 76Br (T½ = 16.2 h; β+ = 57 %) have suitable decay properties for molecular imaging with PET, while the Auger electron emitters 77Br (T½ = 57.0 h) and 80mBr (T½ = 4.4 h) as well as the β−-emitter 82Br (T½ = 35.3 h) are useful for internal radiotherapy. 77Br is additionally suited for SPECT. The isotopes 75Br, 76Br and 77Br are usually produced at a cyclotron either by 3He and α-particle induced reactions on natural arsenic or by proton and deuteron induced reactions on enriched selenium isotopes [1]. As target mate-rials for the latter two reactions, earlier ele-mental selenium [2] and selenides of Cu, Ag, Mn, Mo, Cr, Ti, Pb and Sn were investigated [cf. 3–7].
Besides several wet chemical separation techniques the dry distillation of bromine from the irradiated targets was investigated, too [cf. 2, 4, 5]. However, the method needs further development.
Nickel selenide was investigated as a promising target to withstand high beam currents, and the dry distillation technique for the isolation of n.c.a. radiobromine from the target was optimized.
Material and Methods
Crystalline Nickel-(II) selenide (0.3–0.5 g) was melted into a 0.5 mm deep cavity of a 1 mm thick Ni plate covered with a Ni grid. NiSe has a melting point of 959 °C. For development of targeting and the chemical separation, natural target material was used. Irradiations of NiSe were usually performed with protons of 17 MeV using a slanting water cooled target holder at the cyclotron BC1710 [8]. For radiochemical studies a beam current of 3 µA and a beam time of about 1 h were appropriate.
To separate the produced no-carrier-added (n.c.a.) radiobromine from the target material a dry distillation method was chosen. The apparatus was developed on the basis of a dry distillation method for iodine [cf. 9,10] and optimized to obtain the bromine as n.c.a. [*Br]bromide in a small volume of sodium hydroxide solution.
Changing different components of the apparatus, the dead volume could be minimized and an almost constant argon flow as carrier medium was realized. Various capillaries of platinum, stainless steel and quartz glass with different diameters and lengths were tested to trap the radiobromine.
Results and Conclusion
Nickel selenide proved successful as target material for the production of radiobromine by proton irradiation with 17 MeV protons. The target was tested so far only at beam currents up to 10 µA, but further investigations are ongoing.
The optimized dry distillation procedure allows trapping of 80–90 % of the produced radiobromine in a capillary. For this purpose quartz glass capillaries proved to be most suitable. After rinsing the capillary with 0.1 M NaOH solution the activity can be nearly completely obtained in less than 100 µL solution as [*Br]bromide immediately useable for radiosynthesis. So, the overall separation yield was estimated to 81 ± 5 %.
The radionuclidic composition and activity of the separated radiobromide was measured by γ-ray spectrometry. Due to the use of natural selenium the determination of the isotopic purity was not meaningful, but it could be shown that the radiobromine was free from other radioisotopes co-produced in the target material and the backing. The radiochemical purity as well as the specific activity were determined by radio ionchromatography.
Further experiments using NiSe produced from nickel and enriched selenium are to be per-formed. The isotopic purity of the produced respective radiobromide, the production yield at high beam currents and the reusability of the target material have to be studied
Evolution of production of Astatine-211 in Orléans Cyclotron
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Since 2005, we produce, at academic scale in Orleans, 211At for needs of chemistry and physicians teams of Nantes in research project of alpha immunotherapy. Between 2005 and 2014, several modifications were been made on preparation of target, targetry and radiation to protect personnel.
Material and Methods
The first target was a molten Bi metal onto a Cu support pre-treated with acid attack. The wished thickness (up to 100 µm) was obtained by mechanical treatment of target. The target is irradiated at 32MeV alpha particle beam for around 2 hours and then delivered by road transport to users. Only a measure of radiation dose was made to evaluate target activity. The second target we have used since 2010 is a electrodeposition of Bi (thickness of around 30 µm) onto AlN backing. We used a beam of 30.5 MeV for reaction 207Bi(α,n)211At (2 h with a current intensity of 2µA). Activity has measured with a detector Ge at 687 keV (γ-branching fraction = 0.26 %) before to be delivered. For all targets, beam energy on target was around 28.7 Mev in order not to produce too much 210At.
Results and Conclusion
138 productions with the first target were delivered with an estimated activity of less than 100 MBq. Difficulties with wet extraction1, low yield of radiolabelling (metallic impurities and activation of copper resulting in 66Ga and 67Ga) made necessary to change process of extraction. With support of AlN, dry extraction was used with good yield (75–80 %) and without activation of support. Until today, 46 batchs were delivered with activity of 44 ± 12 MBq/µAh. Yield activity of 211At has been almost doubled compared to first target (25MBq/µAh). The dose burden to personnel was decreased with modification of targetry (outside of blockhouse of cyclotron, in a specific line beam to radionuclide production, cf. FIG. 1).
In the case of 211At production, energy of reaction is of major impact. With our versatile accelerator (range of energy in alpha between 10 and 50 MeV) and a low thickness of metal, it’s easy to reach the right energy. This radionuclide production will be continued until ARRONAX, Nantes cyclotron, could take over from us for bigger activity of 211At
Hydrolytically stable Titanium-45
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Titanium-45, a candidate PET isotope, is under-employed largely because of the challenging aqueous chemistry of Ti(IV). The propensity for hydrolysis of Ti(IV) compounds makes radio-labeling difficult and excludes 45Ti from use in bio-conjugate chemistry. This is unfortunate because the physical characteristics are extremely desirable: 45Ti has a 3 hour half-life, a positron branching ratio of 85 %, a low Eβmax of 1.04 MeV, and negligible secondary gamma emission. In terms of isotope production, 45Ti is transmuted from naturally mono-isotopic 45Sc by low energy proton irradiation. The high cross-section and production rates on an unenriched metal foil target contribute to make 45Ti an ideal PET radionuclide.
In order to bring 45Ti to even a preclinical plat-form, the hydrolytic instability of aqueous Ti(IV) needs to be addressed. Recently, the groups of Edit Tshuva (Hebrew University of Jerusalem) and Thomas Huhn (University of Konstanz) have synthesized several stable Ti(IV) compounds based upon the salan ligand [1,2]. Additionally, these compounds have shown heightened cyto-toxicity against HT-29 (human colorectal cancer) cells, amongst others, as compared to traditional metal-based chemotherapeutics such as cisplatin.
The aim of our work has been to produce the radioactive analogue of one of these Ti(IV)-salan compounds, Ti-salan-dipic [2], which has hydro-lytic stability on the order of weeks. Not only will this allow us to shed some light on the still un-known mechanism of antiproliferative action of titanium-based chemotherapeutics, but it will also make progress toward bioconjugate 45Ti PET tracers.
In the current abstract, we present some of the methods we are using to separate 45Ti from irradiated Sc, and subsequent labeling conditions.
Material and Methods
45Ti was produced by proton irradiation of 250μm scandium foils at currents ranging from 10-20μA on a GE PETTrace. In order to increase production rate in the thin foil, an 800μm aluminum degrader was used to take the proton energy down from the nominal 16 MeV. The scandium was cooled by contact to a water-cooled silver plate.
The activated foil was dissolved in 4M HCl, dried under argon at 120 oC, and taken back up in 12M HCl. Here, four (i-iv below) different approaches to removing the Ti from the Sc and labeling were taken with varying success.
Briefly: i. 45Ti was separated on hydroxamate resin, as presented by K. Gagnon [3], only at 12M acid concentration followed by on-column radiolabeling. ii. 45Ti was extracted into 1-octanol [4], stripped with 12M HCl, and used directly for labeling from the organic phase. iii. 45Ti was trapped on a C-18 cartridge that had been pre-loaded with 1-octanol, similar to ion-pairing, and eluted with isopropanol. iv. 45Ti was extracted onto a polystyrene based 1,3 diol resin (RAPP polymers) and labeling commenced on the column.
Radiolabeling was slightly different in each condition, but in general the salan and dipic ligands were added to the 45Ti in pyridine and reacted at elevated temperature (60–100 oC) for several (10–30) minutes. Reaction progression and radiochemical purity were assessed with silica TLC in chloroform : ethyl acetate (1 : 1).
Results and Conclusion
The trap, release, and yields for the four methods listed above are shown in TABLE 1. The best result was with the 1,3 diol resin which had the added advantage of reacting on-column.
Further optimization is underway including a test of a solid supported 1,2 diol, and preclinical imaging with HT-29 xenografts.
We conclude that hydrolytically stable 45Ti com-pounds can be synthesized in high yield, and hope that this advances the radiochemistry and use of 45Ti toward more widespread applications
Production and novel radiochemical separation of 194Au from Pt for use in multi-modality nanoparticles: Production and novel radiochemical separation of 194Au from Pt for use in multi-modality nanoparticles
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Gold nanoparticles (AuNPs) have demonstrated their incredible versatility in applications such as in vitro and in vivo imaging, cancer therapy, and drug delivery.[1-3] These AuNPs come in many shapes including nanospheres, nanorods, nanoshells, and nanocages. Their versatility stems from the ability to construct or label a single AuNP with many functions. Many types of AuNPs are inherently flourescent, allowing for ex vivo utilization as well as small animal fluorescence imaging.[4] High atomic number and physical density allow for the possibility of using AuNPs as computed tomography (CT) contrast agents, especially in dual energy applications.[5]
Some attempts have been made to bring AuNPs into the realm of nuclear medicine, mostly involving the extrinsic labeling of chelated radio-metals. Although these strategies have brought some success, an intrinsic labeling strategy could reduce concerns of in vivo instability, and changes in pharmacokinetic behavior.[6] Intrinsic radiolabeling strategies involve synthesizing the nanoparticles in the presence of a gold radioisotope, which is thereby structurally incorporated. The isotope of choice for this technique has typically been 198Au (t½ = 2.7 d, Eγ = 411.8 keV) as it is reactor produced and commercially available. However with such a high energy gamma ray, SPECT aquisition is far from optimal.
Motivated by the shortcomings of previous intrinsic labeling techniques, we have sought to develop 194Au (t½ = 1.48 d, β+ = 1.73 %) as a potential PET isotope for labeling AuNPs. Although this nuclide has a weak positron branching ratio, it also has prominent gamma ray energies of 328 and 294 keV which are closer to the optimal SPECT energy window, allowing for the ability to image with both PET and SPECT.
Material and Methods
194Au was produced by natPt(p,x) using 16 MeV protons. Target construction consisted of a water jet cooled platinum disc.
Following irradiation, targets were etched by fresh concentrated aqua regia at 80 °C for four hours. The resulting solution was diluted by a factor of four and loaded onto a 50 mg UTEVA (Eichrom extraction resin) column equilibrated by 1 M HNO3. The column was rinsed with 10 mL 1 M HNO3, and the product was eluted using concentrated HNO3 in less than 1 mL.
Results and Conclusion
End of bombardment (EOB) yield for 194Au was measured to be 0.134 mCi/μAh by high purity germanium analysis. The half life was measured to be 38.5 ± 2.8 hours, which agrees well with the true half life of 37.92 hours. In addition to the production of 194Au, the production of 190–193Au and 196Au was observed. Most notably, the EOB yield for 193Au (t½ = 17.7 h) was 0.189 mCi/μAh.
Target dissolution was slow and incomplete after four hours of etching. Alternative dissolution strategies i.e. electrolytic dissolution may be needed moving forward. The separation of 194Au from bulk Pt via the UTEVA extraction resin was robust and efficient, with an average separation efficiency of 96 %. An extensive literature review revealed no other Au/Pt separation from solutions containing aqua regia. Future goals include synthesis of ultrasmall 194Au incorporated AuNPs using a facile thermal reduction method.PET, CT and fluorescence imaging will also be carried out in vivo to establish the multimodal capabilities of the intrinsically radio-labeled nanoplatforms.
To conclude, a novel separation technique has been developed to separate 194Au from Pt for use in intrinsically radiolabeled multi-modal AuNPs
Metallic impurities in the Cu-fraction of Ni targets prepared from NiCl2 solutions
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Copper-64 is an emerging radionuclide with applications in PET molecular imaging and/or internal therapy and it is typically produced by proton irradiation of isotopically enriched 64Ni electrodeposited on a suitable backing substrate. We recently reported a simple and efficient method for the preparation of nickel targets from electrolytic solutions of nickel chloride and boric acid [1]. Herein we report our recent research work on the analysis of metallic impurities in the copper-fraction of the radiochemical separation process.
Material and Methods
Nickel targets were prepared and processed as previously reported [1]. Briefly, the bath solution was composed of a mixture of natural NiCl2. 6H2O (135 mg/ml) and H3BO3 (15 mg/ml) and Ni was electrodeposited using a gold disk as cathode and a platinum wire as anode. The plating process was carried out at room temperature using 2 ml of bath solution (pH = 3.7) and a constant current density of 60 mA/cm2 for 1 hour. The unirradiated Ni targets were dissolved in 1–2 ml of concentrated (10M) HCl at 90 oC. After complete dissolution of the Ni layer, water was added to dilute the acid to 6M, and the solution was transferred onto a chromatographic column containing AG 1-X8 resin equilibrated with 6M HCl. The Ni , Co and Cu isotopes were separated by using the well-known chromatography of the chloro-complexes. The sample-fractions containing the Cu isotopes (15 ml, 0.1M HCl) were collected in plastic centrifuge tubes previously soaked in 1M HNO3 and rinsed with Milli-Q water (18 MΩ cm). Impurities of B, Co, Ni, Cu and Zn in these samples were determined by inductively coupled plasma-mass spectroscopy (ICP-MS) at the Department of Geosciences (Laboratory of Isotopic Studies) of the National University.
Results and Conclusions
The mass of Ni deposited in 1 h was 25.0 ± 1.0 mg (n = 3) and the current efficiency was > 75 % in all cases. The pH of the electrolytic solution tended to decrease along the electrodeposition process (3.71.6). The results of ICP-MS analysis of the Cu-fractions from the cold chromatography separation runs are shown in FIG. 1. We were particularly interested in the boron impurities as H3BO3 is used as buffer for electrodeposition of the Ni targets.
Except for the Ni impurities that were deter-mined to be in the range of ppm (mg/l), all other analyzed metallic impurities were found to be in the range of ppb (µg/l), including boron. The Co, Ni, Cu and Zn impurities determined in the Cu-fraction in this work using Ni targets electrode-posited from a NiCl2 acidic solution, are in the same order of magnitude compared with that obtained when using targets prepared from an alkaline solution [2], with the advantage of the simplicity of the electrodeposition method from NiCl2 solutions, as the target material is already recovered in the chemical form of NiCl2, enabling a simpler, one step process to prepare a new plating solution when using enriched 64Ni target material for the production of 64Cu
Bulk liquid-metal irradiation system
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Low melting point metals are often encapsulated in a hermetic container, irradiated and the container transferred to hot-cell for material removal and processing. An important process of this kind is the production of 82Sr from rubidium (melting point: 39.5 °C.)
This new concept departures completely form the encapsulated targets approach and allows an almost continues production by the irradiation of the bulk metal. As well, eliminated is the target transfer. By placing the target material dissolution chamber right in the target station, only the dissolution product is pumped to the hotcell for further processing.
Material and Methods
Some of the disadvantages of the encapsulated target are:
1. Complicated transfer system that is ex-pensive to install, slow and prone to failures.
2. Complex and expensive encapsulation procedure.
3. Loss of production time during the lengthy target changing.
4. Capsule geometry is constrained by the encapsulating process and transfer demands compromising heat transfer and beam power.
To avoid the difficulties of liquid metal handling, metal salts are often used instead (rubidium chloride is one example). This creates other problems and limits the beam currents and production yields.
In the system described, the liquid metal is transferred (by gravity) from a bulk container to an irradiation chamber. The chamber, made out of nickel-plated silver, holds the correct quantity of rubidium for one irradiation run. Because of the geometry of the chamber and the efficient cooling, up to 40KW of beam power can be delivered to the target. The chamber is equipped with thermocouples and a liquid-metal level detector and is entirely of welded/brazed construction. The alloy foil that forms the beam window is electron-beam welded to the chamber front ring.
At the end of irradiation the irradiated liquid metal is gravity fed into a reaction chamber situ-ated below the irradiation chamber, and a new load of fresh rubidium released into the irradia-tion chamber. The liquid-metal transfer and the irradiation components are shown on FIG. 1, and the sectional view on FIG. 2.
Appropriate chemicals (n-butanol in the case of rubidium) are delivered to the reaction chamber and the irradiated metal dissolved. The liquid dissolution product is transferred back to the hotcell. Since all steps of the reaction involve liquids, only small diameter tubes connect the target station with the hotcell. The transfer is fast and simple.
The bulk liquid-metal storage container can be constructed to hold enough material for 10 or more runs. When empty, it is replaced with a pre-loaded one. The container is connected to the target system with one coupling and the exchange takes a short time. A robotic bottle exchange can be implemented if desired.
The station is equipped with its own vacuum system, beam diagnostic (consisting of a four-sector mask) and a collimation. The target chamber and each of the beam intercepting components are electrically insulated to allow beam current monitoring.
Constructed entirely out of metal and ceramic the target core assembly does not suffer from radiation damage. The use of aluminum, silver and alumina reduce component activation.
Results and Conclusion
A large part of the station design is based on the well proven construction of high current solid target system and is using the same, or similar components.
Test was performed to optimize the liquid-metal transfer and the chamber filling with the correct volume, while leaving some room for expansion.
A process for niobium coating of sliver is investi-gated. Niobium is known to provide good corro-sion resistance against liquid metals.
Thermal modelling of the target and flow analysis of the cooling geometry is under way