Radiochemical separation of <Superscript>67</Superscript>Ga from Zn and Cu using the adsorbent resin Amberlite XAD-7

Summary A column chromatography was developed using a nonionic, polar adsorbent resin Amberlite XAD-7 for routine radiochemical separation of ⁶⁷Ga from Zn and Cu. 7M HCl was found to be the optimum concentration for adsorption of ⁶⁷Ga and elution of both Zn and Cu. Distilled water was used for elution of ⁶⁷Ga from the column. The optimum flow rate of the solutions on to the column was 5 ml/min. The concentrations of Zn and Cu in the final product (30 ml), measured by a polarography method, were 2.0 and 0.5 ppm, respectively. A solvent extraction method, using diisopropyl ether/7M HCl system was also tested to perform the same separation and a comparison between the two methods was made. The radiochemical chromatographic separation system is simple in design and easy to operate for separations in a hot cell.     

Development of [<Superscript>64</Superscript>Cu]-DOTA-anti-CD20 for targeted therapy

Copper-64 was produced as a by-product of ⁵⁵Co via ⁶⁴Ni(p,n)⁶⁴Cu by 15 MeV proton bombardment of ^natNi resulting in a thick target yield of 5.31 MBq/μAh (143.5 μCi/μAh) and a radiochemical separation yield of 95% (radionuclide purity >97% after 25 hours of bombardment). Rituximab was successively labeled with [⁶⁴Cu]-CuCl₂. N-succinimidyl-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA-NHS) was prepared at 25 °C using DOTA and N-hydroxy succinimide (NHS) in CH₂Cl₂ followed by the addition of 1 ml of a Rituximab pharmaceutical solution. Radiolabeling was performed at 37 °C in 3 hours. Radio thin-layer chromatography showed an overall radiochemical purity of 90–95% at optimized conditions (specific activity=30 GBq/mg, labeling efficacy; 82%) using various chromatography systems. The final isotonic ⁶⁴Cu-DOTA-Rituximab complex was passed through a 0.22 μm filter and checked by gel electrophoresis for radiolysis control. Stability of the final product was checked in the formulation and in presence of human serum at 37 °C.     

Preparation of <Superscript>26</Superscript>Al, <Superscript>59</Superscript>Ni, <Superscript>44</Superscript>Ti, <Superscript>53</Superscript>Mn and <Superscript>60</Superscript>Fe from a proton irradiated copper beam dump

The station for pions cancer therapy was operated at PSI from 1980 to 1992. After a cooling time of 12 years it’s made of copper beam dump was cut and samples were taken for analytical purposes. The sampling collected about 500 g of high active copper chips that can be used for separation of exotic radionuclides. The analyses by gamma spectrometry, LSC and AMS showed main nuclides present to be ⁶⁰Co, ⁵⁴Mn, ²²Na, ⁶⁵Zn, ²⁶Al, ⁵³Mn, ⁵⁹Ni, ⁶³Ni, ⁵⁵Fe and ⁶⁰Fe and ⁴⁴Ti with a daughter nuclide ⁴⁴Sc. In the frame of ERAWAST project a procedure combining selective precipitation and ion exchange for the separation of the rare radionuclides from the copper beam dump was developed. The proposed separation procedure is easy for remote controlled implementation in a hot cell. The ion exchange separation of Ni, Al, Mg, Ti and Fe was complete and high decontamination factors for copper and cobalt were achieved. Based on the developed procedure a remotely controlled system for separation of exotic radionuclides from the copper chips was set up. The full scale system was installed in a hot cell where high activity levels can be handled. In order to evaluate the reliability and functionality of the system extensive tests have been done. During the test period 13.86 g in total of the proton irradiated copper beam dump were processed for separation of ²⁶Al, ⁵⁹Ni, ⁵³Mn, ⁴⁴Ti and ⁶⁰Fe. The results showed that the system was operational and the radionuclide separation was selective with high chemical yield. The procedure manages as well the generated liquid wastes containing high level of ⁶⁰Co activity.     

Characterisation of a Cu selective resin and its application to the production of <Superscript>64</Superscript>Cu

Cu isotopes (e.g. ⁶⁴Cu) increasingly find use in radiopharmaceutical applications, accordingly fast and reliable methods for the production of these isotopes are needed. The aim of the presented project is the characterization of a Cu selective extraction chromatographic resin for the fast and selective separation of Cu radionuclides, e.g. from irradiated targets. The characterisation of the resin includes the determination of weight distribution factors D _w of Cu, Ni, Zn and other potentially interfering elements and impurities for varying acids and pH values, the influence of macro amounts of Ni and Zn on the extraction of Cu as well as the influence of other potential interferents. Based on the obtained results, a method for the separation of Cu and its purification from irradiated Ni or Zn targets was developed and tested on simulated Ni and Zn targets.     

Thin-target excitation functions and optimization of simultaneous production of NCA copper-64 and gallium-66,67 by deuteron induced nuclear reactions on a natural zinc target

Copper-64 is a radionuclide suitable for labeling of a wide range of radiopharmaceuticals for PET imaging, as well as systemic or local radioimmunotherapy of tumors. Among the possible methods for cyclotron production of No Carrier Added (NCA) ⁶⁴Cu (⁶¹Cu), we investigated the deuteron irradiation on natural Zn target, via (d,axn) and (d,2pxn) nuclear reactions. This paper reports the preliminary results about the experimental determination and theoretical calculation of thin-target excitation functions in the energy range up to 19 MeV for ⁶¹Cu, ⁶⁴Cu, ⁶⁶Ga, ⁶⁷Ga, ⁶⁵Zn and ^69mZn. A fast selective radiochemical separation of NCA ⁶⁴Cu from Zn target and Ga radionuclides, with quality control tests is described too.     

Separation of radioarsenic from irradiated germanium oxide targets for the production of <Superscript>71</Superscript>As and <Superscript>72</Superscript>As

A method for the separation of no-carrier-added arsenic radionuclides from the bulk amount of proton-irradiated GeO₂ targets as well as from coproduced radiogallium was developed. The radionuclides ⁶⁹Ge and ⁶⁷Ga produced during irradiation of GeO₂ were used as tracers for Ge and Ga in the experiments. After dissolution of the target the ratio of As(III) to As(V) was determined via thin layer chromatography (TLC). The extraction of radioarsenic by different organic solvents from acid solutions containing alkali iodide was studied and optimized. The influence of the concentration of various acids (HCl, HClO₄, HNO₃, HBr, H₂SO₄) as well as of KI was studied using cyclohexane. The optimum separation of radioarsenic was achieved using cyclohexane with 4.75 M HCl and 0.5 M KI and its back-extraction with a 0.1% H₂O₂ solution. The separation leads to high purity radioarsenic containing no radiogallium and <0.001% [⁶⁹Ge]Ge. The overall radiochemical yield is 93 ± 3%. The practical application of the optimized procedure in the production of ⁷¹As and ⁷²As is demonstrated and batch yields achieved were in the range of 75–84% of the theoretical values.     

Preparation and characterization of nickel targets for cyclotron production of <Superscript>64</Superscript>Cu

The reaction route ⁶⁴Ni(p,n)⁶⁴Cu is very popular for the preparation of ⁶⁴Cu because its entrance channel is accessible at low energies and yield of reaction is quite high. However, a high price of the enriched ⁶⁴Ni is a disadvantage of this reaction path; hence, preparation of a chemically pure nickel targets for the production of ⁶⁴Cu using COSTIS (Compact Solid Target Irradiation System) is of a great research interest. In this paper, effects of boric acid, composition of electrolytic bath and electrochemical process conditions on the quality of nickel films deposited on 2 mm thick gold or platinum disc targets were investigated. Chemical purity of the electrodeposited nickel was measured by the Auger electron spectroscopy and the surface quality of targets was studied microscopically (SEM). A SRIM program was used for a thickness target calculation. COSTIS target station was installed at the end of the external beam line of the IBA Cyclone 18/9 cyclotron, and the irradiation surface of target was optimized. The target station has been equipped with Al or Nb window foil in the front of the target to degrade the beam energy to an optimal value.     

PC-controlled radiochemistry system for preparation of no-carrier-added <Superscript>64</Superscript>Cu

A homemade modular PC-controlled radiochemistry system, used to separate the no-carrier-added ⁶⁴Cu from irradiated electroplated solid cyclotron target ⁶⁴Ni layers was developed. Upon irradiation the target layer is dissolved in diluted nitric acid, followed by selective liquid–liquid extraction of the ⁶⁴Cu. The final purification step is achieved making use of an anion exchange column. All the separation procedures can be carried out remote-controlled with a minimized risk of operator errors within less than 2 h, with excellent processing yield (>95%).     

Sequential separation of <Superscript>99</Superscript>Tc, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe, <Superscript>90</Superscript>Sr and <Superscript>59/63</Superscript>Ni from radioactive wastes

A sequential separation procedure has been developed for the determination of ⁹⁹Tc, ⁹⁴Nb, ⁵⁵Fe, ⁹⁰Sr and ^59/63Ni in various radioactive wastes generated from nuclear power plants. Ion exchange and extraction chromatography were adopted for individual separation of the radionuclides. Precipitation was supplementarily utilized for both purification of the individual radionuclides and preparation of the radionuclide sources for use in a radioactivity measurement. The chromatographic separation behavior of the radionuclides both from the sample matrix metals and from one another was investigated using stable metals, Re (as a surrogate of ⁹⁹Tc), Nb, Fe, Sr and Ni. The validity of the procedure for reliability and applicability was evaluated by measuring the recovery of the metal carriers added to synthetic radioactive waste solutions. The recoveries by the chromatographic separation were in the range of 84.8 to 102.2% with 2s of less than 8.6%, the recoveries by the precipitation being in the range of 84.3 to 97.3% with 2s of less than 10.9%.     

Radiopharmaceutical Studies and Nuclear Medicine; Simultaneous <Superscript>68</Superscript>Ge and <Superscript>88</Superscript>Zr recovery from proton irradiated Ga/Nb capsules (LA-UR #03-2319)

Summary Germanium-68 (270.8 d, EC 100%) is the parent nuclide of ⁶⁸Ga, a β+ emitter important to positron emission tomography (PET). ⁶⁸Ge is obtained by a (p,x) induced nuclear reaction on natural Ga. A typical Ga target assembly consists of liquid Ga contained in a Nb capsule, since Nb is one of the few metals resistant to liquid Ga. Zirconium-88 (83.4 d, via ⁹³Nb(p, α2n)) is one longer-lived radioisotope generated by the proton irradiation of naturally mono-isotopic ⁹³Nb. It decays into ⁸⁸Y, which, in turn, has been considered a useful radiolabel surrogate for ⁹⁰Y in the investigation of radiolabeled compounds for cancer radioimmunotherapy. This paper introduces a wet chemical procedure for the processing of Nb/Ga target capsules and the simultaneous recovery of ⁶⁸Ge and ⁸⁸Zr.     

Production possibility of <Superscript>61</Superscript>Cu using proton induced nuclear reactions on zinc for PET studies

Summary Excitation functions were measured by stacked-foil technique for the ^natZn(p,x)⁶¹Cu, ⁶⁶Zn(p,x)⁶¹Cu, ⁶⁸Zn(p,x)⁶¹Cu and ^natZn(p,x)⁶⁰Cu nuclear processes up to 100 MeV. The experimental cross sections were compared with published data. On the base of these excitation functions, the cross sections of ⁶⁴Zn(p,x)⁶¹Cu process were also deduced. Integral thick target yields were calculated for the ⁶⁴Zn(p,x)⁶¹Cu and ^nat,64Zn(p,x)⁶⁰Cu processes and irradiation parameters were elaborated for the ⁶¹Cu production via the ⁶⁴Zn+p reactions for low and middle energy accelerators. According to our calculations the yield of ⁶¹Cu amounts to 1.02 ^. 10¹¹ Bq ^. A^{-1 .} s^-1 (9.9 mCi ^.µ A^{-1 .} h^-1) from 19®10 MeV while it reaches 3.91 ^. 10¹¹ Bq ^. A^{-1 .} s^-1 (38 mCi ^{. µ} A^{-1 .} h^-1) in the energy range of 67®60 MeV.     

Production of no-carrier-added64Cu and67Cu in a reactor

Fission neutron spectrum averaged cross-sections were measured for the reactions⁶⁴Zn(n, p)⁶⁴Cu,⁶⁷Zn(n, p)⁶⁷Cu and⁶⁸Zn(n, α)⁶⁵Ni by the activation technique, using radiochemical separations and γ-spectroscopy. The preparation of⁶⁴Cu and⁶⁷Cu in a nuclear reactor was studied. The^{64, 67}Cu was separated from zinc matrix activity using anion exchange column technique.     

<Superscript>47</Superscript>Sc production development by cyclotron irradiation of <Superscript>48</Superscript>Ca

The therapeutic radionuclide ⁴⁷Sc was produced through the ⁴⁸Ca(p,2n) channel on a proton beam accelerator. The obtained results show that the optimum proton energies are in the range of 24–17 MeV, giving the possibility to produce ⁴⁷Sc radionuclide containing 7.4% of ⁴⁸Sc. After activation, the powdery CaCO₃ target material was dissolved in HCl and scandium isotopes were isolated from the targets. The performed separation experiments indicate that, due to the simplicity of the operations and the chemical purity of the obtained ⁴⁷Sc the best separation process is when scandium radioisotopes are separated on the 0.2 µm filter.     

Preparation of <Superscript>166</Superscript>Dy/<Superscript>166</Superscript>Ho-chitosan as an in vivo generator for radiosynovectomy

¹⁶⁶Ho is one of the most effective radionuclides used for radiosynovectomy. One method to deliver this radioisotope to target tissue is via the ¹⁶⁶Dy/¹⁶⁶Ho in vivo generator system. The aim of this work was to prepare ¹⁶⁶Dy/¹⁶⁶Ho-chitosan (¹⁶⁶Dy/¹⁶⁶Ho-CHIT) in vivo generator for radiosynovectomy applications. ¹⁶⁶Dy obtained by the irradiation of natural ¹⁶⁴Dy target. ¹⁶⁶Dy was separated from ¹⁶⁶Ho by extraction chromatographic method (separation yield; 93% and separation factor;1.7). Chitosan labeling was performed in acetic acid with 99.3 ± 0.6% radiochemical purity. Biodistribution studies on intraarticular injected rats demonstrated high retention in the knee joint even 7 days showing no radioactivity leakage from the injection site into other organs as well as any translocation of the daughter nucleus after β^? decay of ¹⁶⁶Dy.     

Systematic radiochemical separation for the determination of <Superscript>99</Superscript>Tc, <Superscript>90</Superscript>Sr, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe and <Superscript>59,63</Superscript>Ni in low and intermediate radioactive waste samples

A simple and rapid separation procedure was systemized for the determination of ⁹⁹Tc, ⁹⁰Sr, ⁹⁴Nb, ⁵⁵Fe and ^59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of ⁹⁹Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of ^134,137Cs and ^110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.     

Production of 67Cu via the 68Zn(p,2p)67Cu reaction and recovery of 68Zn target

The radionuclide ⁶⁷Cu was produced via the ⁶⁸Zn(p,2p)⁶⁷Cu reaction by irradiating enriched ⁶⁸Zn targets with 70 MeV proton beam. Copper-67 was chemically separated from the zinc target by ion-exchange chromatography using Chelex-100 chelating ion-exchange resin. Procedure for recovery of the enriched ⁶⁸Zn was developed. The target recovery yield of this method was evaluated to be more than 97%.     

Combined determination of <Superscript>99</Superscript>Tc and <Superscript>108m</Superscript>Ag in L/ILW liquid wastes

The low- and intermediate-activity level liquid wastes produced by the Paks Nuclear Power Plant (NPP) contain routinely measureable gamma-emitting (e.g., ⁵⁴Mn, ⁶⁰Co, ^110mAg, and ¹³⁷Cs) as well as many so-called “difficult-to-measure” radionuclides. Despite of their low specific activity compared to the total, the reliable determination of these radionuclides is an important issue of nuclear waste management. The increasing amount of waste samples to be qualified yearly by our laboratory put a pressure on revising the existing procedure of ⁹⁹Tc separation applied. We have managed to halve the initial amount of the sample required to achieve the same level of detection of technetium. Furthermore, one of the new purifying steps introduced have proved to be able to separate ^108mAg (and ^110mAg) better than 99% keeping the ⁹⁹Tc content of the product almost intact. Means of separation of ⁹⁹Tc from ¹⁰⁶Ru and ^124+125Sb have also been successfully investigated. As intended, this new procedure has a major impact on the chemical reagent as well as the electricity requirement of the separation making it more cost-effective.     

Radiochemical studies relevant to the separation of 68Ga and 68Ge

The radiochemical separation of radiogallium from radiogermanium was studied using ion-exchange chromatography (Amberlite IR-120) and solvent extraction (Aliquat 336 in o-xylene). Both Amberlite IR-120 and Aliquat 336 in o-xylene have been used for the first time in separations involving radiogallium and radiogermanium. For tracer studies the radionuclides ⁶⁸Ge (t _1/2 = 270.8 days), ⁶⁹Ge (t _1/2 = 39 h) and ⁶⁷Ga (t _1/2 = 78.3 h) were used. They were produced by the nuclear reactions ^natGa(p,xn)^68,69Ge and ^natZn(p,xn)⁶⁷Ga, respectively, and separated from their target materials in no-carrier-added form. Several factors affecting the separation of radiogallium from radiogermanium were studied and for each procedure the optimum conditions were determined. The solvent extraction using Aliquat 336 was found to be better. The separation yield of radiogallium was >95%, the time of separation short, the contamination from radiogermanium <0.008% and the final product was obtained in 0.5 M KOH. This method was adapted to the separation of n.c.a. ⁶⁸Ga from its parent n.c.a. ⁶⁸Ge. The quality of the product thus obtained is discussed.     

Preparation of [<Superscript>66</Superscript>Ga]bleomycin complex as a possible PET radiopharmaceutical

Summary Gallium-66 (T_1/2 = 9.49 h) is an interesting radionuclide that has potential use for positron emission tomography (PET) imaging of biological processes in intermediate to slow target tissue uptake. ⁶⁶Ga was produced in the NRCAM cyclotron as a result of 15 MeV proton bombardment of ⁶⁶Zn with current intensity of 180 μA (yield 11.2 mCi/μAh). Bleomycin (BLM) was labeled with [⁶⁶Ga]GaCl₃     

Production cross section of <Superscript>236</Superscript>Th in the interaction of <Superscript>238</Superscript>U with 60 MeV/u <Superscript>18</Superscript>O ions

The neutron-rich target-like isotope ²³⁶Th has been produced in the ²³⁸U-2p multinucleon transfer reaction between a 60 MeV/u ¹⁸O beam and natural ²³⁸U targets. The activities of thorium were determined after radiochemical separation of Th from the mixture of uranium and reaction products. The ²³⁶Th isotope was identified by the characteristic γ-rays of 642.2, 687.6 and 229.6 keV. The production cross section of ²³⁶Th was determined to be 250±50 μb.