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Excitation functions of natZn(p,x) nuclear reactions with proton beam energy below 18 MeV
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We measured the excitation functions of natZn (p,x) reactions up to 17.6 MeV using the stacked-foils activation technique. High-purity natural zinc (and copper) foils were irradiated with proton beams from an 18MeV medical cyclotron, the predominant purpose of which is to provide a routine regional service for clinical PET radiopharmaceuticals. Thick-target integral yields were also deduced from the measured excitation functions of the produced radioisotopes. These results were compared with the literature and were found to be in good agreement with most but not all published reports.
Material and Methods
The excitation functions of the natZn(p,x) reactions were measured by the well-known stacked foil technique (1). High purity zinc foils (99.99%; Goodfellow Metals Ltd., UK) each thickness 0.025 ± 0.003 mm with isotopic composition 64Zn (48.6 %), 66Zn (27.9 %), 67Zn (4.1 %), 68Zn (18.8 %) and 70Zn (0.6 %) were loaded into a solid targetry system on a 300-mm external beam line utilising helium-gas and chilled water to cool the target body (2). A typical foils stack consisted of repeated units of four Zn foils interleaved with a high purity copper foil (0.025 ± 0.004 mm); the latter for monitoring beam flux using the well documented 63,65Cu(p,n)63,65Zn reactions. Foil stacks were irradiated with a primary beam of energy 17.6 MeV, accounting for beam degradation by an obligatory 0.0250 ± 0.0005 mm-thick Havar® foil beam-line vacuum window. Irradiation was for 3 min at a beam current of 5 µA. Activated foils were measured using cryo high-purity Ge γ-spectroscopy to quantify the product radionuclides 61Cu, 66Ga, 67Ga and 65Zn. Radioactivity of each isotope was corrected to end of bombardment (EOB).
Results and Conclusion
New cross-sectional data for natZn(p,x) reactions up to 17.6 MeV yielding 61Cu, 66Ga, 67Ga and 65Zn isotopes were measured in independent replicated (N = 3) experiments. Results were generally in good agreement with published data.
These isotopes can potentially be used in clinical or preclinical studies, following appropriate chemical separations of the zinc, gallium and copper (3). The FIG. 1 shows thick-target integral yields calculated from excitation functions measured in this study.
It can be calculated (for example) that useful activities of 61Cu can be produced using a 100 µm thick natZn target in a beam provided by a standard medium-energy medical cyclotron. For example, an irradiation at 40 µA for 2 hr at 17.6 MeV would produce approximately 1.7 GBq of 61Cu at EOB. Such currents are readily achievable using solid targetry in our laboratory (2)
High power targets for cyclotron production of 99mTc‡
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Technetium-99m, supplied in the form of 99Mo/99mTc generators, is the most widely used radioisotope for nuclear medical imaging. The parent isotope 99Mo is currently produced in nuclear reactors. Recent disruptions in the 99Mo supply chain [1] prompted the development of methods for the direct accelerator-based production of 99mTc.
Our approach involves the 100Mo(p,2n)99mTc reaction on isotopically enriched molybdenum using small medical cyclotrons (Ep ≤ 20 MeV), which is a viable method for the production of clinically useful quantities of 99mTc [2].
Multi-Curie production of 99mTc requires a 100Mo target capable of dissipating high beam intensities [3]. We have reported the fabrication of 100Mo targets of both small and large area tar-gets by electrophoretic deposition and subsequent sintering [4]. As part of our efforts to further enhance the performance of molybdenum targets at high beam currents, we have developed a novel target system (initially de-signed for the GE PETtrace cyclotron) based on a pressed and sintered 100Mo plate brazed onto a dispersion-strengthened copper backing.
Materials and Methods
In the first step, a molybdenum plate is produced similarly to the method described in [5] by compacting approximately 1.5 g of commercially available 100Mo powder using a cylindrical tool of 20 mm diameter. A pressure between 25 kN/cm2 and 250 kN/cm2 is applied by means of a hydraulic press.
The pressed molybdenum plate is then sintered in a reducing atmosphere (Ar/2% H2) at 1,700 oC for five hours. The resulting 100Mo plates have about 90–95 % of the molybdenum bulk density.
The 100Mo plate is furnace brazed at ~750 oC onto a backing manufactured from a disperse on strengthened copper composite (e.g. Glidcop AL-15) using a high temperature silver-copper brazing filler.
This process yields a unique, mechanically and thermally robust target system for high beam power irradiation.
Irradiations were performed on the GE PETtrace cyclotrons at LHRI and CPDC with 16.5 MeV protons and beam currents ≥ 100 µA. Targets were visually inspected after a 6 hour, 130 µA bombardment (2.73 kW/cm2, average) and were found fully intact. Up to 4.7 Ci of 99mTc have been produced to date. The saturated production yield remained constant between 2 hour and 6 hour irradiations.
Results and Conclusion
These results demonstrate that our brazed tar-get assembly can withstand high beam intensities for long irradiations without deterioration. Efforts are currently underway to determine maximum performance parameters
Temperature model verification and beam characterization on a solid target system
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Temperature modeling using Finite Element Analysis (FEA) is widely used by particle beam-line designers as a useful tool to determine the thermal performance of an irradiated target system. A comparison study was performed between FEA calculated temperatures on platinum with experimental results using direct thermocouple measurements. The aims are to determine the best beam model for future solid target design, determine the maximum target current for different target materials and the temperature tolerance for any modification to our existing solid targetry system.
Material and Methods
The theoretical temperature of the target sys-tem was determined using SolidWorks 2013 with Flow Simulation Analysis (FSA) module. The FSA module determines the maximum temperature inside the target material given the global conditions (material specification, flow rates, boundary conditions, etc.) for a given target current. The proton beam was modeled as a volumetric heat source inside the target material based on the distribution of energy loss in the material along the beam axis. The method used by Comor, et al1 was used in this study. The method segmented the target material into five individual layers, each layer being 50 m thick. The energy lost per layer was calculated using SRIM3 and converted into the power lost per layer. A thickness of 250 μm of platinum completely stops the impinging proton beam at 11.5 MeV with the highest deposition of power per layer corresponding to the Bragg peak.
The target material used in the simulation reflects the physical target disk used for temperature measurements (platinum, dia. 25.0 mm, thickness 2.0 mm) with two K-type thermocouples (dia. 0.5 mm, stainless steel sheath) embedded in the platinum disk. One thermocouple is located in the geometric center, while the other is located at a radial position 8 mm from center. The outer thermocouple is to determine the peripheral temperature near the o-ring seal. Temperature was maintained below the melting point for the material (Viton®, melting point 220 °C) during the irradiation to ensure the integrity of the water cooling system.
The solid targetry system used in this study is an in-house built, significantly modified version2 of a published design1. The solid target system is mounted onto an 18/18MeV IBA Cyclotron with dual ion source, on a 300mm beam-line with no internal optics or steering magnets. A graphite collimator reduces the beam to 10mm in diameter and a degrader is used to reduce the proton beam energy to 11.5 MeV, considered suitable for production of radiometal PET isotopes 89Zr and 64Cu. Temperature was measured with and without the 300 mm beam-line to compare the effects of beam divergence on the solid target (FIGS. 1 and 2).
The experiment was conducted using both H− ion sources with different ion-to-puller extraction gaps (ion source 1 is 1.55 mm with ion source 2 at 1.90 mm). The setting of the ion-to-puller gap changes the focusing of the accelerated beam inside the cavity.
Results and Discussion
The segmented beam model was used to calculate the temperature on and within the target, as well as the maximum temperature of the bulk material. The first segment is the leading segment of the material irradiated by the incident proton beam. Results are shown in TABLE 2.
Target temperatures were measured experimentally under two different conditions; target attached at the end of a 300mm beam-line and target attached directly to the cyclotron.
The temperature was measured experimentally using the platinum disk with 2 thermocouples inside the bulk target material irradiated on the end of a 300mm beam-line. The measured temperature is shown in TABLE 2.
The variation between ion source 1 and 2 for the temperature measured in the center was 11–15 %, while the variation on the radial position was 2–6 %. A smaller ion-to-puller extraction distance (ion source 1) reduces the cross-sectional area of the accelerated beam; the consequent high proton current density (10mm diameter collimated beam) increases the temperature inside the bulk material for a fixed target current. The highest observed radial temperature was 93 °C, with target current of 50 μA using ion source 1. This is well below the melting point for the o-ring seal.
The temperature measured experimentally using the same platinum disk with no beam-line is shown in TABLE 4.
A temperature difference of up to 7 % was measured between ion source 1 and 2 at the exit port without the beam-line, while the maximum variation on the radial position was 3 %.
A comparison between the calculated theoretical and measured temperatures is shown in FIGS. 3 to 6. The temperatures calculated by the FEA model underestimate the temperature regardless of target position (with or without the beam-line) and for both ion sources.
The temperature difference between the FEA model and the experimental results increases with increasing target currents. As shown in Figure 3, at the target center the FEA model underestimated the temperature by 22–32 % for ion source 1 and 13–22 % for ion source 2. This is consistent with the difference between the two ion sources due to the difference in the ion-to-puller gap size.
With the target mounted at the exit port the theoretical and measured temperature for the center of the platinum disk is shown in FIGURE 4.
The FEA model underestimates the temperature at the center of the platinum disc by 2–10 % for both ion sources. As shown with the previous experiment, the margin of error increases with increasing target current.
Comparison between FIGS. 3 and 4 shows the measured temperature at the center of the platinum disk is significantly lower when the target is attached to exit port of the cyclotron. Localised area of high current density (hot spots) is not registered as higher temperature in the bulk material. True temperature inside the bulk material is highly dependent on the thermal conductivity of the target material and the resolution of the thermocouple. The cross-sectional area of the beam ‘hot-spot’ will be greater due to beam divergence at the end of the beam line compared with the exit port. The ‘hot’ area of the expanded beam becomes a significant portion of the overall collimated beam (collimator dia. 10.0 mm). A more uniform beam profile (less heterogeneity) evenly distributed the area of high current density across the disk surface, effectively increasing the temperature of the bulk material while decreasing the sensitivity required to measure the true temperature. As observed from this comparative study it appears that a more homogeneous current density leads to a higher temperature measurement at the target center.
With the solid target at the end of the beam-line, target current lost on the collimator and beam-line was >55%. The effect of beam divergence is clearly observed in TABLE 5.
With the target mounted directly at the exit port the current lost was reduced to < 40 %. Although the average proton current density is the same for any set target current, irrespective of target position, the contribution of the peripheral beam to the total target current should not be underestimated. A loss of ~40 μA on the collimator and beam-line places greater reliance on the center of the ‘hot’ beam to maintain the same target current.
The temperature at the radial position (FIG. 5) observes the same trend as for the temperature measured in the center.
The error increases for higher target currents and the FEA model underestimated the temperature by 19–40 %. The error at this location is due partly to the model’s assumption of a uniform heat source, applied to the material on a single axis (perpendicular to the material surface) and does not account for any scattering or divergence of the incident proton beam.
FIGURE 6 shows that the FEA model underestimated the radial temperature by 16–37 %, when the target is connected to the exit port, for reasons discussed previously. Comparison with FIG. 5 (target on the beam-line) shows the same margin of error between the FEA and the experimental results (19–40 %).
The temperature difference between the FEA model and measured temperature at the radial position is independent of the beam profile and beam divergence. The FEA model underestimated the temperature at the radial location with or without the beam-line and for both ion sources. The significance difference in temperature between the FEA model and the experimental is due to our model assumption that the maximum radial temperature is on the irradiated surface and not inside the material corresponding to the layer with the maximum energy lost. In addition, the FEA model does not ac-count for the divergence of the proton beam as it travels through the material. Given the temperature at 50 μA target current is > 90 °C (TABLES 3 and 4) we have capped the experi-ment below this point to prevent any damage the o-ring seal.
Conclusion
The segmented FEA model was inadequate in determining the temperature for the target at the end of a 300mm beam-line (> 30 % difference). A combination of beam divergence and greater uniform coverage of high current density beam resulted in a higher than predicted temperature reading. However, the segmented FEA model provides a good estimation (< 10 % difference) for the observed temperature of the bulk material at the exit port. The simplistic FEA model was unable estimate the temperature at the radial position (~ 40 % difference) regardless of ion source or target position.
A comparison between the two ion sources with different ion-to-puller extraction gap, leading to different focusing of the accelerated beam yield minimal temperature difference. Although a 15% difference was observed between the ion sources at the end of the beam-line, a major contributing factor is beam divergence beyond the magnetic field rather than the beam size of the accelerated beam.
Further studies are underway to determine the beam profile (quantitatively using radiographic film), quantify the contribution of the peripheral beam to the total beam current by comparing different size collimators and to investigate other FEA models by applying different beam models (heterogeneous and homogeneous beam) and different heat sources (surface vs. volumetric).
Currently the RAPID Lab solid targetry is placed at the end of the beam-line for easy loading and unloading, since multiple target irradiations are performed per month2. However, RAPID is presently developing a new solid targetry sys-tem which eliminates the need for a beam-line and will be able to manage a maximum extracted target current of 150 μA
Ion exchange trap and release of [C-11]CO2
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Recently in our laboratory we needed a reliable and relatively simple source of aqueous solutions of [11C]CO2. We examined various methods of trapping [11C]CO2 gas both in solution and on ion exchange resins, followed by elution into aqueous phase. We favor simple methods that have high trapping and elution efficiencies and produce a highly concentrated solution. Furthermore, we desired methods that would minimize the use of hazardous reagents and materials with respect to both handling and disposal. We also considered the formulation of the final solution in terms of chemical compatibility with contacted materials, working with the assumption that dilute bicarbonate or carbonate solutions will have little reactivity with many materials. In a phantom, compatibility with materials (i.e. plastics, glues, metals, etc.) is important (1-4), while in (bio)geochemical studies – where transport of carbon is important – the chemical form of the radiolabelled molecule is important, but compatibility must be determined on a case-by-case basis (5-7).
Small medical cyclotrons can easily produce carbon-11 as gaseous [11C]CO2, and various methods are utilized to incorporate carbon-11 into solution, often with unfavorable resource requirements, costs, or chemical properties. Commonly [11C]CO2 gas is bubbled through a strong base, forming the carbonate anion; but neutralizing a strong base (as to avoid special handling or disposal requirements) requires a large volume of diluent or buffer; or a very precise addition of acid – which if done improperly – may lead to an acidic pH and subsequent loss of [11C]CO2 from solution (8,9). Alternatively, [11C]CO2 (or [11C]CH4) can be converted to [11C]CH3I at high-yield, but requires specialized, expensive radio-synthesis equipment (10-12). [11C]CH3I can then be trapped in DMSO (albeit providing a volatile and hazardous solution) or used as a synthon en route to water soluble compounds such as [11C]choline (13). Finally, leftover radiolabelled radiopharmaceuticals from a carbon-11 imaging experiment could be used, but chemical compatibility (i.e. lipophilicity) of the radiolabelled compound may be of concern.
Carbon dioxide gas will dissolve with a solubility of 1.5 g/L at STP (9) and slowly react with water to generate carbonic acid (H2CO3), a weak acid.
Passing [11C]CO2 through a base-activated ion exchange cartridge, the [11C]CO2 reacts with hydroxide ions to form [11C]carbonate which is bound to the resin due to its higher selectivity for carbonate than hydroxide (14). Elution with excess bicarbonate displaces [11C]carbonate and neutralizes any remaining hydroxide, providing a 11C aqueous solution that is mildly basic, chemically non-hazardous, and very concentrated.
Material and Methods
[11C]CO2 gas trapping efficiency was evaluated for solutions and base-activated ion exchange resins. The gas was delivered either rapidly in a high-flow bolus directly from the cyclotron target or slowly in a low-flow helium stream during heating of a carbosieves column. Elution efficiency of ion exchange cartridges were evaluated for both fraction of trapped activity eluted and volume of solution needed for elution.
[11C]CO2 was produced via the 14N(p,α)11C reaction on a CTI RDS111 – 11 MeV cyclotron at the Lawrence Berkeley National Laboratory’s Bio-medical Isotope Facility. The 7 mL target is pressurized to 315 psi with 1% O2/N2 gas, equating to 150 mL gas at STP. For direct-from-target trapping experiments, the target was decompressed and routed to the cartridge via 50 feet of 0.020” I.D. tubing until the target falls to atmospheric pressure (~55 seconds) providing an inhomogeneous flow – a short rapid burst of flow followed by a longer low-flow bleed. For helium-eluted experiments, the [11C]CO2 was unloaded from the cyclotron target and trapped on a room-temperature carbosieves column (15). Target gases were subsequently flushed from the column for 30 seconds with helium at 50 mL/min. After heating the column to 125 °C without gas flow, [11C]CO2 was eluted off the column in helium at 15 mL/min.
[11C]CO2/He was bubbled through 9 aqueous and 2 organic solutions to test for trapping efficiency in a slow, steady helium stream at 15 mL/min (sodium hydroxide (0.96M, 0.096M, 0.0096M), sodium bicarbonate (1.14M, 0.57M, 0.057M), sodium carbonate (2.0M, 1.0M, 0.10M), ethanol, and DMSO (2mL ea.). An Ascarite-filled cartridge was attached to trap any untrapped [11C]CO2. Measures of radioactivity were made using a Capintec CRC-15R dose calibrator. Trapping efficiency for solutions is calculated as the fraction of radioactivity captured in solution relative to the sum of the solution and the Ascarite trap.
Three different commercially available, ion ex-change cartridges were evaluated for trapping and elution efficiencies. FIGURE 1 shows a photo-graphic comparison of the physical size and shapes of the cartridges as well as a X-ray computed tomography (CT) cross sectional view of the internal ion exchange resin volume and dead volume of the cartridges.
All cartridges were activated with 1 mL of 1 N aqueous NaOH followed by passing 10 mL deionized water then 10 mL of air through the cartridge.
In both direct-from-target (n = 4) and helium-stream experiments (n = 3 or 4), cartridges were connected to [11C]CO2 delivery lines via Luer connections. The gas exiting the cartridge passed through an empty 3 mL crimp-top vial as a liquid trap en route to an Ascarite trap on the vent needle as described above. Trapping efficiency for cartridges is calculated as the fraction of radioactivity captured on the cartridge relative to the sum of the cartridge, the empty vial, and the Ascarite trap.
Cartridges were eluted with 0.5 mL of saturated sodium bicarbonate solution (1.14 M @ 20°C) followed by 9.5 mL water and 10 mL air. Elution efficiency is calculated as the fraction of radioactivity eluted in 10 mL relative to the sum of the spent cartridge following elution and the 10 mL eluate (Equation 5). The pH of the eluate was measured using 0-14 pH test strips.
Results and Conclusion
The trapping of [11C]CO2 in all solutions was less than 70% of the total radioactivity with the exception of the 0.96 M and 0.096 M NaOH. With a higher concentration of base driving equilibrium towards carbonate stability, it could be expected that the most basic solution had the best trapping efficiency, but this attribute also means it is least desirable solution to work with from a hazardous material or chemical compatibility perspective.
When [11C]CO2 was delivered in a helium stream, all three cartridges performed at near 100% efficiency, as shown in FIGURE 4. With higher flow, direct-from-target delivery, the cartridges trapped [11C]CO2 with a wider range of efficiencies: ICOH (99 ± 1 %), ORTG (90 ± 2 %), and QMA (79 ± 4 %). Elution resulted in > 99 % release of carbon-11 activity for both QMA and ORTG cartridges, but only 39 ± 3 % release from the ICOH cartridge. Elution efficiency of the trapped radioactivity (Equation 5) was independent of the method of [11C]CO2 delivery. Across all cartridges and delivery methods, the eluate was at about pH = 10.
We recommend that the ORTG cartridge be used for trapping of [11C]CO2 gas with elution by > 300 µL of saturated bicarbonate solution. This recommendation is based on the better trapping for ORTG cartridges compared to the QMA cartridges in the direct-from-target [11C]CO2 delivery method and the smaller volume needed for elution of all trapped carbon. This method excels based on its simplicity, adaptability to automation, low-cost ($5/cartridge), and observations that a single ORTG cartridge suffers no loss of performance after multiple uses. A potential disadvantage to this method is that it involves using a carbon-containing eluent, which means that this method cannot be used for imaging experiments that require high specific activity. However, considering the eluate is a mildly basic aqueous solution, we expect that it will be compatible with a wide variety of materials and experimental applications
A practical high current 11 MeV production of high specific activity 89Zr
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Zr-89 is a useful radionuclide for radiolabeling proteins and other molecules.1,2 There are many reports of cyclotron production of 89Zr by the 89Y (p,n) reaction. Most irradiations use thin metal backed deposits of Y and irradiation currents up to 100 µA or thicker amounts of Y or Y2O3 with
~ 20 µA irradiations.3,4 We are working to develop high specific activity 89Zr using a low energy 11 MeV cyclotron. We have found that target Y metal contains carrier Zr and higher specific activities are achieved with less Y. The goal of this work was to optimize yield while minimizing the amount of Y that was irradiated.
Material and Methods
All irradiations were done using a Siemens Eclipse 11 MeV proton cyclotron. Y foils were used for the experiments described here. Y2O3 was tried and abandoned due to lower yield and poor heat transfer. Yttrium metal foils from Alfa Aesar, ESPI Metals and Sigma Aldrich, 0.1 to
1 mm in thickness, were tested. Each foil was irradiated for 10 to 15 minutes.
The targets to hold the Y foils were made of aluminum and were designed to fit within the “paper burn” unit of the Siemen’s Eclipse target station, allowing the Y target body to be easily inserted and removed from the system. Several Al targets of 2 cm diam. and 7.6 cm long were tested with the face of the targets from 11, 26 or 90o relative to the beam to vary watts cm−2 on the foil. The front of the foils was cooled by He convection and the foil backs by conduction to the Al target body. The target body was cooled by conduction to the water cooled Al sleeve of the target holder.
Results and Conclusion
The best target was two stacked, 0.25 mm thick, foils to stop beam. 92% of the 89Zr activity was in the front 0.25 mm Y foil. With the greatest slant we could irradiate up to 30 µA of beam on tar-get. However, the 13×30 mm dimensions of the foil was more mass (0.41 g) and lower specific activity than was desired. Redesign of the target gave a target 90o to the beam with 12×12 mm foils (0.15 g/foil) that were undamaged with up to 30 µA irradiation when two foils were used. This design has a reduction in beam at the edges of ~10%. With this design, a single Y foil, 0.25 mm thick sustained over 31 µA of beam and a peak power on target of 270 watts cm−2. The product was radionuclidically pure 89Zr after all 89mZr and small amounts of 13N produced from oxygen at the surface had decayed (TABLE 1).
Our conclusion is that the optimum target is a single 0.25 mm thick Y foil to obtain the greatest specific activity at this proton energy. This produces 167 MBq of 89Zr at EOB with a 15 minute and 31 µA irradiation. We are continuing to redesign the clamp design to reduce losses at the edge of the beam
Simulation studies for the in-vivo dose verification of particle therapy
Get PDFAn increasing number of cancer patients is treated with proton beams or other light ion beams which allow to deliver dose precisely to the tumor. However, the depth dose distribution of these particles, which enables this precision, is sensitive to deviations from the treatment plan, as e.g. anatomical changes. Thus, to assure the quality of the treatment, a non-invasive in-vivo dose verification is highly desired. This monitoring of particle therapy relies on the detection of secondary radiation which is produced by interactions between the beam particles and the nuclei of the patient’s tissue.
Up to now, the only clinically applied method for in-vivo dosimetry is Positron Emission Tomography which makes use of the beta+-activity produced during the irradiation (PT-PET). Since from a PT-PET measurement the applied dose cannot be directly deduced, the simulated distribution of beta+-emitting nuclei is used as a basis for the analysis of the measured PT-PET data. Therefore, the reliable modeling of the production rates and the spatial distribution of the beta+-emitters is required. PT-PET applied during instead of after the treatment is referred to as in-beam PET. A challenge concerning in-beam PET is the design of the PET camera, because a standard full-ring scanner is not feasible. For instance, a double-head PET camera is applicable, but low count rates and the limited solid angle coverage can compromise the image quality. For this reason, a detector system which provides a time resolution allowing the incorporation of time-of-flight information (TOF) into the iterative reconstruction algorithm is desired to improve the quality of the reconstructed images.
Secondly, Prompt Gamma Imaging (PGI), a technique based on the detection of prompt gamma-rays, is currently pursued. Concerning the emissions of prompt gamma-rays during particle irradiation, experimental data is not sufficiently available, making simulations necessary. Compton cameras are based on the detection of incoherently scattered photons and are investigated with respect to PGI. Monte Carlo simulations serve for the optimization of the camera design and the evaluation of criteria for the selection of measured events.
Thus, for in-beam PET and PGI dedicated detection systems and, moreover, profound knowledge about the corresponding radiation fields are required. Using various simulation codes, this thesis contributes to the modelling of the beta+-emitters and photons produced during particle irradiation, as well as to the evaluation and optimization of hardware for both techniques.
Concerning the modeling of the production of the relevant beta+-emitters, the abilities of the Monte Carlo simulation code PHITS and of the deterministic, one-dimensional code HIBRAC were assessed. The Monte Carlo tool GEANT4 was applied for an additional comparison. For irradiations with protons, helium, lithium, and carbon, the depth-dependent yields of the simulated beta+-emitters were compared to experimental data. In general, PHITS underestimated the yields of the considered beta+-emitters in contrast to GEANT4 which provided acceptable values. HIBRAC was substantially extended to enable the modeling of the depth-dependent yields of specific nuclides. For proton beams and carbon ion beams HIBRAC can compete with GEANT4 for this application. Since HIBRAC is fast, compact, and easy to modify, it could be a basis for the simulations of the beta+-emitters in clinical application. PHITS was also applied to the modeling of prompt gamma-rays during proton irradiation following an experimental setup. From this study, it can be concluded that PHITS could be an alternative to GEANT4 in this context.
Another aim was the optimization of Compton camera prototypes. GEANT4 simulations were carried out with the focus on detection probabilities and the rate of valid events. Based on the results, the feasibility of a Compton camera setup consisting of a CZT detector and an LSO or BGO detector was confirmed. Several recommendations concerning the design and arrangement of the Compton camera prototype were derived. Furthermore, several promising event selection strategies were evaluated. The GEANT4 simulations were validated by comparing simulated to measured energy depositions in the detector layers. This comparison also led to the reconsideration of the efficiency of the prototype. A further study evaluated if electron-positron pairs resulting from pair productions could be detected with the existing prototype in addition to Compton events. Regarding the efficiency and the achievable angular resolution, the successful application of the considered prototype as pair production camera to the monitoring of particle therapy is questionable.
Finally, the application of a PET camera consisting of Resistive Plate Chambers (RPCs) providing a good time resolution to in-beam PET was discussed. A scintillator-based PET camera based on a commercially available scanner was used as reference. This evaluation included simulations of the detector response, the image reconstructions using various procedures, and the analysis of image quality. Realistic activity distributions based on real treatment plans for carbon ion therapy were used. The low efficiency of the RPC-based PET camera led to images of poor quality. Neither visually nor with the semi-automatic tool YaPET a reliable detectability of range deviations was possible. The incorporation of TOF into the iterative reconstruction algorithm was especially advantageous for the considered RPC-based PET camera in terms of convergence and artifacts.
The application of the real-time capable back projection method Direct TOF for the RPCbased PET camera resulted in an image quality comparable to the one achieved with the iterative algorihms. In total, this study does not indicate the further investigation of RPC-based PET cameras with similar efficiency for in-beam PET application.
To sum up, simulation studies were performed aimed at the progress of in-vivo dosimetry. Regarding the modeling of the beta+-emitter production and prompt gamma-ray emissions, different simulation codes were evaluated. HIBRAC could be a basis for clinical PT-PET simulations, however, a detailed validation of the underlying cross section models is required. Several recommendations for the optimization of a Compton Camera prototype resulted from systematic variations of the setup. Nevertheless, the definite evaluation of the feasibility of a Compton camera for PGI can only be performed by further experiments. For PT-PET, the efficiency of the detector system is the crucial factor. Due to the obtained results for the considered RPC-based PET camera, the focus should be kept to scintillator-based PET cameras for this purpose
Partikeltherapie-PET – Optimierung der Datenverarbeitung für die klinische Anwendung
Get PDFDie Strahlentherapie ist einer der drei Partner im interdisziplinären Feld der Onkologie. In Europa, Asien und den USA besteht zunehmend die Möglichkeit einer Therapie mit Strahlen aus geladenen Ionen anstelle von Photonen. Eine Anlage in Dresden befindet sich in der Kommissionierungsphase.
Die Ionenstrahltherapie bietet den Vorteil einer sehr konformalen Behandlung des Tumorvolumens durch die endliche Reichweite der Strahlen und ein ausgeprägtes Dosismaximum kurz vor dem Ende des Strahlpfades. Da eine Therapie in der Regel über bis zu 30 Sitzungen an verschiedenen Tagen durchgeführt wird und der Strahlweg stark von dem durchdrungenen Gewebe beeinflusst wird, sind Verfahren für eine in vivo Verifikation der Strahlapplikation wünschenswert. Eine dieser Methoden ist die Partikeltherapie–Positronen-Emissions-Tomografie (PT-PET). Sie beruht auf der Messung der vom Therapiestrahl erzeugten β+-Aktivitätsverteilung. Da eine direkte Berechnung der Dosis aus der Aktivität in lebendem Gewebe nicht möglich ist, wird die gemessene Aktivitätsverteilung mit einer berechneten Vorhersage verglichen und anschließend entschieden, ob die nächste Therapiesitzung wie geplant erfolgen kann oder Anpassungen notwendig sind. Die vorliegende Arbeit beschäftigt sich mit drei Themen aus dem Bereich der Datenverarbeitung für die PT-PET.
Im ersten Teil wird ein Algorithmus zur Bestimmung von Reichweitendifferenzen aus zwei β+- Aktivitätsverteilungen adaptiert und evaluiert. Dies geschieht zunächst anhand einer Simulationsstudie mit realen Patientendaten. Ein Ansatz für eine automatisierte Analyse der Daten lieferte keine zufriedenstellenden Ergebnisse. Daher wird ein Software-Prototyp für eine semiautomatische, assistierte Datenanalyse entwickelt. Die Evaluierung erfolgt durch Experimente mit Phantomen am 12C-Strahl.
Die erzeugte Aktivitätsverteilung wird von physiologischen Prozessen im Organismus beeinflusst.
Dies führt zu einer Entfernung von Emittern vom Ort ihrer Erzeugung und damit zu einer Verringerung der diagnostischen Wertigkeit der erfassten Verteilung. Zur Quantifizierung dieses als Washout bezeichneten Effektes existiert ein am Tierexperiment gewonnenes Modell. Dieses Modell wird im zweiten Teil der Arbeit auf reale Patientendaten angewendet. Es konnte gezeigt werden, dass das Modell grundsätzlich anwendbar ist und für die betrachtete Tumorlokalisation Schädelbasis ein Washout mit einer Halbwertszeit von (155,7±4,6) s existiert.
Die Berechnung der Vorhersage der β+-Aktivitätsverteilung kann durch übliche Monte-Carlo-Verfahren erfolgen. Dabei werden die Wechselwirkungsquerschnitte zahlreicher Reaktionskanäle benötigt. Als alternatives Verfahren wurde die Verwendung gemessener Ausbeuten (Yields) radioaktiver Nuklide in verschiedenen Referenzmaterialien vorgeschlagen. Auf Basis einer vorhandenen Datenbank dieser Yields und einer existierenden Condensed-History-Monte-Carlo-Simulation wird ein Programm zur Berechnung von Aktivitätsverteilungen auf Yieldbasis entwickelt. Mit der Methode kann die β+-Aktivitätsverteilung in Phantomen und Patienten zufriedenstellend vorhergesagt werden.
Die entwickelten Verfahren sollen einen Einsatz der PT-PET im klinischen Umfeld erleichtern und damit einen breiten Einsatz ermöglichen, um das volle Potential der Ionenstrahltherapie nutzbar zu machen
Targetry investigations of 186Re production via proton induced reactions on natural Osmium disulfide and Tungsten disulfide targets
Get PDFIntroduction
Radioisotopes play an important role in nuclear medicine and represent powerful tools for imaging and therapy. With the extensive use of 99mTc-based imaging agents, therapeutic rhenium analogues are highly desirable. Rhenium-186 emits therapeutic − particles with an endpoint-energy of 1.07 MeV, allowing for a small, targeted tissue range of 3.6 mm. Additionally, its low abundance γ-ray emission of 137.2 keV (9.42 %) allows for in vivo tracking of a radiolabeled compounds and dosimetry calculations. With a longer half-life of 3.718 days, synthesis and shipment of Re-186 based radiopharmaceuticals is not limited. Rhenium-186 can be produced either in a reactor or in an accelerator. Currently, Re-186 is produced in a reactor via the 185Re(n,γ) reaction resulting in low specific activity which makes its therapeutic application limited.[1] Production in an accelerator, such as the PETtrace at the University of Missouri Research Reactor (MURR), can theoretically provide a specific activity of 34,600 Ci.mmol−1 Re[2], which represents a 62 fold increase over reactor produced 186Re. The studies reported herein focused on the evaluation of accelerator-based reaction pathways to produce high specific activity (HSA) 186Re. Those pathways include proton and deuteron bombardment of tungsten and osmium targets by the following reactions: 186W(p,n)186Re, 186W(d,2n) 186Re, 189Os(p,α)186Re, and 192Os(p,α3n)186Re. Additional information on target design related to the determination and optimization of production rates, radionuclidic purity, and yield are presented.
Material and Methods
Osmium and tungsten metals are very hard and thus very brittle. Attempts at pressing the pure metal into aluminum backings resulted in chalky targets, which easily crumbled during handling. Osmium disulfide (OsS2) and tungsten disulfide (WS2) were identified to provide a softer, less brittle chemical form for targets. OsS2 and WS2 targets were prepared using a unilateral press with a 13 mm diameter die to form pressed powder discs. A simple target holder design (FIG. 1) was implemented to provide a stabilizing platform for the pressed discs. The target material was sealed in place with epoxy using a thin aluminum foil pressed over the target face. Initial irradiations of OsS2 were performed using the 16 MeV GE PETtrace cyclotron at MURR. Irradiations were performed for 30–60 minutes with proton beam currents of 10–20 µA. Following irradiation, the OsS2 targets were dissolved in NaOCl and the pH adjusted using NaOH. The resultant aqueous solution was mixed with methyl ethyl ketone (MEK), with the lipophilic perrhenate being extracted into the MEK layer and the osmium and iridium remaining in the aqueous layer. The MEK extracts were then passed through an acidic alumina column to remove any remaining osmium and iridium. Determination of rhenium and iridium activities was done by gamma spectroscopy on an HPGe detector. Preliminary irradiations on WS2 targets were performed at MURR with the beam degraded to 14 MeV with a proton beam current of 10 µA for 60 minutes. After irradiation, WS2 was dissolved using 30% H2O2 with gentle heating and counted on an HPGe detector to determine the radio-nuclides produced.
Results and Conclusion
Thin natOsS2 targets were produced, irradiated at 16 MeV for 10 µAh, and analyzed for radiorhenium. Under these irradiation conditions, rhenium isotopes were produced in nanocurie quantities while iridium isotopes were produced in microcurie quantities. Future studies with higher proton energies are planned to increase the production of rhenium and decrease the production of iridium. After optimizing irradiation conditions, enriched 189Os will be used for irradiations to reduce the production of unwanted radionuclides. A liquid-liquid extraction method separated the bulk of the rhenium from the iridium. The majority of the rhenium produced was recovered in the first organic aliquot with little iridium observed while the majority of the iridium and osmium was retained in the first aqueous aliquot. Target production with WS2 was successful. A thin target of natWS2 was produced and irradiated at 14 MeV for 10 µAh. Under these irradiation conditions, several rhenium isotopes were produced in microcurie quantities. Target parameters to maximize 186Re production remain to be determined before enriched 186W targets are used for irradiations to reduce the production of unwanted radionuclides. In conclusion, the potential production routes for accelerator-produced high specific activity 186Re are being evaluated. Cyclotron-based irradiations of natOsS2 targets established the feasibility of producing rhenium via the natOs(p,αxn)Re reaction. Current results indicate higher proton energies are necessary to reduce the production of unwanted iridium isotopes while increasing the production of rhenium isotopes. Preliminary irradiations were performed using the 50.5 MeV Scanditronix MC50 clinical cyclotron at the University of Washington to determine irradiation parameters for future higher energy irradiations (20–30 MeV). A rapid liquid-liquid extraction method isolated rhenium from the bulk of the iridium and osmium following irradiation. Preliminary studies indicate WS2 may also provide a suitable target material to produce 186Re via the (p,n) reaction pathway
Evaluation of Column Separation Methods for Simplification of the Wet Chemistry Approach to Isolation of 211At: Evaluation of Column Separation Methods for Simplification of the Wet Chemistry Approach to Isolation of 211At
Get PDFDifficulties with reproducibility of isolation yields when distilling 211At from irradiated bismuth targets led us to use a “wet chemistry” approach for that process1. The wet chemistry approach has provided 211At isolation yields of ~ 78 % after decay and Bi attenuation corrections2. However, the use of diisopropyl ether (DIPE) in the separation process has made it difficult to reach our goal of automating the 211At isolation. Therefore, we have investigated the use of column materials to simplify the isolation of 211At and remove DIPE from the process. In this investigation we evaluated the use of a strong anion exchange resin (AG1×8), a strong cation exchange resin (AG MP-50) and a polyethylene glycol (PEG)-coated resin for separation of 211At from the bismuth target material.
Anion and cation resins AG1×8 and AG MP-50 were obtained from commercial sources. A PEG-coated resin was prepared by reaction of the Merrifield resin with mPEG-OH 2000 in the pres-ence of tBuOK at 80 °C for 3 days, followed by drying under vacuum. Prior to use of the PEG resin, it was soaked in H2O. Resins (400–800 mg) were loaded into polypropylene columns (Applied Separations, Inc.). Column elution studies were conducted with and without reductants (0.75M FeSO4/1M H2SO4 or Na2S2O5) to determine their effect on capture of 211At. After target dissolution in HNO3 (and in most cases subse-quent removal of HNO3 by distillation and redis-solution of solid in 8M HCl), 211At solution was loaded onto the column, then the column was washed with 2M HCl or H2O to separate the Bi, and finally was eluted with strong base to remove the 211At.
Initial studies were conducted with stable iodine to determine if reductants were effective in the presence of large amounts of bismuth ions. Studies with AG1×8 used 125I to determine if that radiohalogen could be captured and recovered from the column when eluting with boric acid buffers at pH 5.3, 8.0 or 10, or H2O at pH 7. Capture and recovery of 211At was evaluated under the same conditions. Further studies with AG1×8 involved eluting with 4M H2SO4. A limited study with AG MP-50 resin used 1M HCl as eluant. Studies with PEG-coated columns used 2M HCl, 4M HCl, 8M HCl, 16 M HNO3 and 8M HNO3 as initial (capture) eluants. Strong base (0.2, 1 or 12.5 M NaOH; 15M NH4OH) and 3 or 500 mM tetrabutylammonium bromide (TBAB) were evaluated for removal of 211At from the columns tested.
The efficiency for capture of 211At on the AG1×8 column was high (99%) when loading with strong acid, but decreased when using 0.1–0.2M boric acid (69–91 %) buffer. Low 211At capture efficiencies were obtained with AG MP-50 col-umns (15–29%). High 211At capture efficiencies (96–100%) were obtained with PEG-coated resins when loading with 8M HCl or 8M HNO3, irre-spective of whether reductant was in the acid solution.
Four column washings (2 mL of 2M HCl each) were required to remove all Bi prior to elution of 211At. No bismuth was detected in solution from the 4th washing in any of the elutions studied.
Low (< 6%) recovery of 211At from the AG1×8 columns was obtained using the conditions studied. Good (60–79%) recovery of 211At was obtained from PEG-coated resin using 15M NH4OH.
Isolation of the 211At from NH4OH solution was accomplished by distillation. In an initial study 211At distilled before obtaining a dry residue. However, later studies demonstrated that addi-tion of NaOH prior to distillation kept the 211At in the distilling flask.
These studies demonstrated that PEG-coated columns could be used to isolate 211At from HNO3-dissolved bismuth targets with good non-optimized (~60%) overall recovery yields. The studies are continuing with optimization of elu-tion conditions and automation of the process
Making high-value, long-lived isotopes to balance a sustainable radiotracer production facility
Get PDFIntroduction
The embrace of PET by medical clinicians has been reluctant (ΔT ≈ 20 yr) primarily due to the scale of the infrastructure that is needed. The capital cost of a cyclotron (≈ 106 USD) is now dwarfed by the demand for compliance to recent regulatory standards. This is a recurring expense, not only imposing an order-of-magnitude increase in staffing and operating costs, but damping the enthusiasm of researchers recalling the brisk pace of research in earlier days. Now an academic site, with little interest or opportunity to scale up production for wider distribution, is burdened by the new regulatory terrain of good manufacturing practice (GMP), mandated for translational studies that will reach only a few subjects. With our production resources held within a basic science department, the Medical Physics cyclotron facility at the University of Wisconsin has sought a sustainable pathway. We now anchor the operating budget by providing high-value, long-lived radionuclides to off-site users, to buffer the fluctuations of local demand for conventional PET synthons.
Material and Methods: The tools of the trade
The radioisotopes discussed here belong to the 3-d and 4-d sub shell, but are now moving into the rare-earths, with applications ranging from
- targeted molecular imaging agents,
- internal radionuclide therapy using to Auger electron-emitters,
- to basic physics experiments using 163Ho (t1/2 ≈ 4500 yr) to determine the mass of the neutrino.
Rather than focusing on the dozens of radionuclides produced, a number of tools deserve mention, as they support a variety of targets, reactions and products. These will be listed in order (A-G) from cyclotron to extraction to analysis.
A. Two cyclotrons are used, a legacy RDS 112 (#1; 1985) and a GE PETtrace (2009). Neutron and gamma detectors are monitored during the long irradia-tions, signaling any subtle changes in the running conditions. (1). The PET-trace is fitted with a quick-change variable degrader target (2), as well as a beam-line with a 5-port (0 o, ±15 o, ±30 o) vertical switching magnet (3). The downward directed beam ports provide support for solid targets (e.g. Ga, S, Se, Te) that melt at low temperature. The irradiation of gas targets employs a generalized manifold to handle inert gases such as 36Ar for the production of 34mCl, as well as natural Kr and Xe for making Rb and Cs isotopes to act as fission product surrogates. These products are captured on a stainless steel target chamber liner, and rinsed off with warm water. The alkali metals are convenient tracers to study the ion exchange trapping process, pivotal in future 99Mo production from solution reactors (4).
B. The preparation of malleable solid targets employs a 10-ton hydraulic bench press, and a jeweler’s mill to roll out foils from pellets, pressed between Nb foils to avoid contamination.
C. Binary alloys are smelted in a programmable 1600o tube furnace under argon flow (eg. NiGa4). Alternatively, an induction furnace now permits highly localized heating of the binary metal charge, while allowing mechanical agitation during the smelting process.
D. Electroplating onto gold discs is used for various enriched target material or the alloys above where quantitative recovery is essential, or where heat transfer from high beam current is demanding.
E. The separation chemistry, prior to che-lation to targeted molecular imaging agents, is performed in LabView-driven, home-built “black boxes” resident in mini-cells (Radiation Shielding Inc.).
F. Analysis of the targets after irradiation makes use of HPGe spectroscopy for gammas and characteristic X-rays of decay (e.g. rare earths). The elemental constitution of target alloys is deter-mined prior to irradiation by X-ray fluorescence analysis, excited by 109Cd and 241Am sources.
G. Finally, broad-band elemental analysis at the ppb level now makes use of a microwave plasma atomic emission spectrometer (Agilent 4200), to be de-scribed elsewhere in this meeting.
Results and Conclusions
The tools above (A-G) are employed in the pro-duction of the expanded list of radionuclides offered by our cyclotron group to both local and off-site colleagues. The list below is ordered in terms of decreasing use, from regular production for national distribution (64Cu, 89Zr), to weekly inhouse use (44Sc, 66,68Ga, 68,69,71Ge, 72As, 61Cu, 86Y), to infrequent production for multi-site collaborations (163Ho, 95mTc, 206Bi):
Radionuclide Target Employs
64Cu 64Ni/Au A, D, G
89Zr natY A, E, G
44Sc natCa A, B, E, F, G
66, 68Ga Zn/Ag A, B, D, E, F, G
68, 69, 71Ge Ga, GaO2 A, B, C, E,F
72As GeO2 A, B, E, F
52Mn natCr A, E, F, G
76, 81mBr SeO A, E, F
34mCl, Rb, Cs noble gas A, E, F
95mTc,163Ho Mo, Dy A, E, F
TABLE 1. Target materials and processes.
The production of long-lived radionuclides lends itself to crowd-sourcing, with distributed irradia-tion at virtually any site with a suitable accelera-tor and a relaxed beam schedule. A number of unique challenges do arise that don’t appear in the usual production of conventional cyclotron products such as 11C or 18F. Contamination by stable metals, inadvertently introduced by target pressing or beam-induced sputtering from degraders, can cause serious interference downstream limiting effective specific activity. Long-lived manganese isotopes are ubiquitous. And some very high value products are simply not within the reach of small cyclotrons, such as 52Fe and 67Cu, being too far off the line of beta stability.
In conclusion, the research leading to a doctoral degree necessarily must focus on the physics and chemistry of novel radionuclides and tracers. On the other hand, clinical and translational research needs established imaging agents, with little room for innovation within the regulatory constraints. Our experience at Wisconsin has led us to a balancing act, with our routine production of clinical doses countered with our research program to provide high-value radionu-clides for our collaborative work with our basic science colleagues