Developing the 134Ce and 134La pair as companion positron emission tomography diagnostic isotopes for 225Ac and 227Th radiotherapeutics.

Developing targeted α-therapies has the potential to transform how diseases are treated. In these interventions, targeting vectors are labelled with α-emitting radioisotopes that deliver destructive radiation discretely to diseased cells while simultaneously sparing the surrounding healthy tissue. Widespread implementation requires advances in non-invasive imaging technologies that rapidly assay therapeutics. Towards this end, positron emission tomography (PET) imaging has emerged as one of the most informative diagnostic techniques. Unfortunately, many promising α-emitting isotopes such as 225Ac and 227Th are incompatible with PET imaging. Here we overcame this obstacle by developing large-scale (Ci-scale) production and purification methods for 134Ce. Subsequent radiolabelling and in vivo PET imaging experiments in a small animal model demonstrated that 134Ce (and its 134La daughter) could be used as a PET imaging candidate for 225AcIII (with reduced 134CeIII) or 227ThIV (with oxidized 134CeIV). Evaluating these data alongside X-ray absorption spectroscopy results demonstrated how success relied on rigorously controlling the CeIII/CeIV redox couple.

A reverse 230U/226Th radionuclide generator for targeted alpha therapy applications.

PURPOSE: Thorium-226 (half-life 30.6 m) is a radionuclide of interest for use in targeted alpha therapy applications. Due to its short half-life, 226Th must be provided through a radionuclide generator system from its parent 230U (20.8 d). Furthermore, as the half-life of 226Th is very short, it should be provided in a form that is directly amenable to use in biomedical applications. METHODS: A reverse radionuclide generator system was developed employing a DGA extraction chromatography column. A 230U/226Th parent/daughter solution in equilibrium is added to a DGA column in >6 M HCl. The parent 230U is eluted first in 0.1 M HNO3 followed by elution of 226Th in 0.1 M citrate buffer pH 5. RESULTS: Thorium-226 was recovered from the radionuclide generator column with >96% yield. Greater than 99.5% of the 230U parent was isolated for reuse in the generator. Long term evaluation over six weeks demonstrated consistent supply of 226Th with greater than 99.5% radionuclidic purity. The only contaminant found in the final product was 230U (<0.5%). CONCLUSIONS: The reverse radionuclide generator described herein was shown to be a feasible method for providing 226Th in high yield, purity and in a chemical form that is amenable for direct use in biomedical applications.

Extraction chromatography of 225Ac and lanthanides on N,N-dioctyldiglycolamic acid /1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide solvent impregnated resin.

The alpha-emitter 225Ac (t1/2 = 9.92 d) is currently under development for targeted alpha-particle therapy of cancer, and accelerator production of 225Ac via proton irradiation of thorium targets requires robust separations of 225Ac from chemically similar fission product lanthanides. Additionally, the lanthanide elements represent critical components in modern technologies, and radiolanthanides such as 140Nd (t1/2 = 3.37 d) also have potential application in the field of nuclear medicine. The ionic liquid, 1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([Bmim][NTf2]), combined with the diglycolamide extractant, N,N-dioctyldiglycolamic acid (DODGAA), was adsorbed on macroporous resin support to produce a solvent impregnated resin (SIR) that was investigated for separations of 225Ac and lanthanides. The equilibrium distribution coefficients (Kd) of the rare earth elements (Sc(III), Y(III), Ln(III)), 225Ac(III), Th(IV), and U(VI) on the prepared DODGAA/[Bmim][NTf2]-SIR were determined from batch adsorption experiments in HCl and HNO3 media. The DODGAA/[Bmim][NTf2]-SIR exhibited preferential uptake of the heavier lanthanide elements while allowing for the separation of the lighter lanthanides. Column separations utilizing the DODGAA/[Bmim][NTf2]-SIR were effective at separating the lighter lanthanides from each other, and separating 225Ac from a mixture of lanthanides, 213Bi, and 225Ra without the need for additional complexing agents.

A Solid-State Support for Separating Astatine-211 from Bismuth.

Increasing access to the short-lived α-emitting radionuclide astatine-211 (211At) has the potential to advance targeted α-therapeutic treatment of disease and to solve challenges facing the medical community. For example, there are numerous technical needs associated with advancing the use of 211At in targeted α-therapy, e.g., improving 211At chelates, developing more effective 211At targeting, and characterizing in vivo 211At behavior. There is an insufficient understanding of astatine chemistry to support these efforts. The chemistry of astatine is one of the least developed of all elements on the periodic table, owing to its limited supply and short half-life. Increasing access to 211At could help address these issues and advance understanding of 211At chemistry in general. We contribute here an extraction chromatographic processing method that simplifies 211At production in terms of purification. It utilizes the commercially available Pre-Filter resin to rapidly (<1.5 h) isolate 211At from irradiated bismuth targets (Bi decontamination factors ≥876 000), in reasonable yield (68-55%) and in a form that is compatible for subsequent in vivo study. We are excited about the potential of this procedure to address 211At supply and processing/purification problems.

Production of 230Pa by proton irradiation of 232Th at the LANL isotope production facility: Precursor of 230U for targeted alpha therapy.

Uranium-230 (t1/2 = 20.8 d) is an alpha-emitting radionuclide that has potential application in targeted alpha therapy (TAT) of cancer. Its parent isotope 230Pa (t1/2 = 17.4 d), can be produced by proton irradiation of thorium metal targets. Preliminary 230Pa production runs were performed at the Los Alamos National Laboratory Isotope Production Facility (LANL-IPF) using thin thorium metal targets and a proton beam energy of 15-30 MeV, followed by radiochemical separation of 230Pa from the irradiated target matrix. The measured 230Pa production yields were found to exceed the predicted values in most of the experiments that were performed. This data will inform further production efforts for providing 230Pa/230U for clinical trials.

Large-Scale Production of 119mTe and 119Sb for Radiopharmaceutical Applications.

Radionuclides find widespread use in medical technologies for treating and diagnosing disease. Among successful and emerging radiotherapeutics, 119Sb has unique potential in targeted therapeutic applications for low-energy electron-emitting isotopes. Unfortunately, developing 119Sb-based drugs has been slow in comparison to other radionuclides, primarily due to limited accessibility. Herein is a production method that overcomes this challenge and expands the available time for large-scale distribution and use. Our approach exploits high flux and fluence from high-energy proton sources to produce longer lived 119mTe. This parent isotope slowly decays to 119Sb, which in turn provides access to 119Sb for longer time periods (in comparison to direct 119Sb production routes). We contribute the target design, irradiation conditions, and a rapid procedure for isolating the 119mTe/119Sb pair. To guide process development and to understand why the procedure was successful, we characterized the Te/Sb separation using Te and Sb K-edge X-ray absorption spectroscopy. The procedure provides low-volume aqueous solutions that have high 119mTe-and consequently 119Sb-specific activity in a chemically pure form. This procedure has been demonstrated at large-scale (production-sized, Ci quantities), and the product has potential to meet stringent Food and Drug Administration requirements for a 119mTe/119Sb active pharmaceutical ingredient.

Separation of Protactinium Employing Sulfur-Based Extraction Chromatographic Resins.

Protactinium-230 ( t1/2 = 17.4 d) is the parent isotope of 230U ( t1/2 = 20.8 d), a radionuclide of interest for targeted alpha therapy (TAT). Column chromatographic methods have been developed to separate no-carrier-added 230Pa from proton irradiated thorium targets and accompanying fission products. Results reported within demonstrate the use of novel sulfur bearing chromatographic extraction resins for the selective separation of protactinium. The recovery yield of 230Pa was 93 ± 4% employing a R3P═S type commercially available resin and 88 ± 4% employing a DGTA (diglycothioamide) containing custom synthesized extraction chromatographic resin. The radiochemical purity of the recovered 230Pa was measured via high purity germanium γ-ray spectroscopy to be >99.5% with the remaining radioactive contaminant being 95Nb due to its similar chemistry to protactinium. Measured equilibrium distribution coefficients for protactinium, thorium, uranium, niobium, radium, and actinium on both the R3P═S type and the DGTA resin in hydrochloric acid media are reported, to the best of our knowledge, for the first time.

Synthesis and radiometric evaluation of diglycolamide functionalized mesoporous silica for the chromatographic separation of actinides Th, Pa and U.

The separation of Th, Pa, and U is of high importance in many applications including nuclear power, nuclear waste, environmental and geochemistry, nuclear forensics and nuclear medicine. Diglycolamide (DGA)-based resins have shown the ability to separate many elements, however, these resins consist of non-covalent impregnation of the DGA molecules on the resin backbone resulting in co-elution of the extraction molecule during separation cycles, therefore limiting their long-term and repeated use. Covalently binding the DGA molecules onto silica is one way to overcome this issue. Herein, measured equilibrium distribution coefficients of normal extraction chromatographic DGA resin and a covalently bound form (KIT-6-N-DGA sorbent) are reported. Several differences are observed between the two systems, the most significant being observed for uranium, which demonstrated significantly lower sorption behavior on KIT-6-N-DGA. These results indicate that U can effectively be separated from Th and Pa using KIT-6-N-DGA, a task that could not be completed with the use of normal DGA alone.

Chromatographic separation of the theranostic radionuclide 111Ag from a proton irradiated thorium matrix.

Column chromatographic methods have been developed to separate no-carrier-added 111Ag from proton irradiated thorium targets and associated fission products as an ancillary process to an existing 225Ac separation design. Herein we report the separation of 111Ag both prior and subsequent to 225Ac recovery using CL resin, a solvent impregnated resin (SIR) that carries an organic solution of alkyl phosphine sulfides (R3P = S) and alkyl phosphine oxides (R3P = O). The recovery yield of 111Ag was 93 ± 9% with a radiochemical purity of 99.9% (prior) and 87 ± 9% with a radiochemical purity of 99.9% (subsequent to) 225Ac recovery. Both processes were successfully performed with insignificant impacts on 225Ac yields or quality. Measured equilibrium distribution coefficients for silver and ruthenium (a residual contaminant) on CL resin in hydrochloric and nitric acid media are reported, to the best of our knowledge, for the first time. Additionally, measured cross sections for the production of 111Ag and 110mAg for the 232Th(p,f)110m,111Ag reactions are reported within.

Separation of 103Ru from a proton irradiated thorium matrix: A potential source of Auger therapy radionuclide 103mRh.

Ruthenium-103 is the parent isotope of 103mRh (t1/2 56.1 min), an isotope of interest for Auger electron therapy. During the proton irradiation of thorium targets, large amounts of 103Ru are generated through proton induced fission. The development of a two part chemical separation process to isolate 103Ru in high yield and purity from a proton irradiated thorium matrix on an analytical scale is described herein. The first part employed an anion exchange column to remove cationic actinide/lanthanide impurities along with the majority of the transition metal fission products. Secondly, an extraction chromatographic column utilizing diglycolamide functional groups was used to decontaminate 103Ru from the remaining impurities. This method resulted in a final radiochemical yield of 83 ± 5% of 103Ru with a purity of 99.9%. Additionally, measured nuclear reaction cross sections for the formation of 103Ru and 106Ru via the 232Th(p,f)103,106Ru reactions are reported within.

Simultaneous Separation of Actinium and Radium Isotopes from a Proton Irradiated Thorium Matrix.

A new method has been developed for the isolation of 223,224,225Ra, in high yield and purity, from a proton irradiated 232Th matrix. Herein we report an all-aqueous process using multiple solid-supported adsorption steps including a citrate chelation method developed to remove >99.9% of the barium contaminants by activity from the final radium product. A procedure involving the use of three columns in succession was developed, and the separation of 223,224,225Ra from the thorium matrix was obtained with an overall recovery yield of 91 ± 3%, average radiochemical purity of 99.9%, and production yields that correspond to physical yields based on previously measured excitation functions.

Proton-induced production and radiochemical isolation of 44Ti from scandium metal targets for 44Ti/44Sc generator development.

Scandium-44g (half-life 3.97h) shows promise for application in positron emission tomography (PET), due to favorable decay parameters. One of the sources of 44gSc is the 44Ti/44gSc generator, which can conveniently provide this radioisotope on a daily basis at a diagnostic facility. Titanium-44 (half-life 60.0 a), in turn, can be obtained via proton irradiation of scandium metal targets. A substantial 44Ti product batch, however, requires high beam currents, long irradiation times and an elaborate chemical procedure for 44Ti isolation and purification. This study describes the production of a combined 175MBq (4.7mCi) batch yield of 44Ti in week long proton irradiations at the Los Alamos Isotope Production Facility (LANL-IPF) and the Brookhaven Linac Isotope Producer (BNL-BLIP). A two-step ion exchange chromatography based chemical separation method is introduced: first, a coarse separation of 44Ti via anion exchange sorption in concentrated HCl results in a 44Tc/Sc separation factor of 102-103. A second, cation exchange based step in HCl media is then applied for 44Ti fine purification from residual Sc mass. In summary, this method yields a 90-97% 44Ti recovery with an overall Ti/Sc separation factor of ≥106.

Bulk production and evaluation of high specific activity 186gRe for cancer therapy using enriched 186WO3 targets in a proton beam.

INTRODUCTION: Rhenium-186g (t1/2 = 3.72 d) is a ß- emitting isotope suitable for theranostic applications. Current production methods rely on reactor production by way of the reaction 185Re(n,γ)186gRe, which results in low specific activities limiting its use for cancer therapy. Production via charged particle activation of enriched 186W results in a 186gRe product with a higher specific activity, allowing it to be used more broadly for targeted radiotherapy applications. This targets the unmet clinical need for more efficient radiotherapeutics. METHODS: A target consisting of highly enriched, pressed 186WO3 was irradiated with protons at the Los Alamos National Laboratory Isotope Production Facility (LANL-IPF) to evaluate 186gRe product yield and quality. LANL-IPF was operated in a dedicated nominal 40 MeV mode. Alkaline dissolution followed by anion exchange chromatography was used to isolate 186gRe from the target material. Phantom and radiolabeling studies were conducted with the produced 186gRe activity. RESULTS: A 186gRe batch yield of 1.38 ± 0.09 MBq/µAh or 384.9 ± 27.3 MBq/C was obtained after 16.5 h in a 205 µA average/230µA maximum current proton beam. The chemical recovery yield was 93% and radiolabeling was achieved with efficiencies ranging from 60-80%. True specific activity of 186gRe at EOB was determined via ICP-AES and amounted to 0.788 ± 0.089 GBq/µg (0.146 ± 0.017 GBq/nmol), which is approximately seven times higher than the product obtained from neutron capture in a reactor. Phantom studies show similar imaging quality to the gold standard 99mTc. CONCLUSIONS: We report a preliminary study of the large-scale production and novel anion exchange based chemical recovery of high specific activity 186gRe from enriched 186WO3 targets in a high-intensity proton beam with exceptional chemical recovery and radiochemical purity.