An electrochemical generator for the continual supply of 213Bi from 225Ac for use in targeted alpha therapy applications.

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225Ac. Radionuclide generators employ longer lived "parent" radionuclides to routinely supply shorter-lived "daughter" radionuclides. The traditional 225Ac/213Bi radiochemical generator relies on an organic cation exchange resin where 225Ac binds to the resin and 213Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1 × 106 Gy/mg), which has been observed when the loading activity of 225Ac exceeds 2.59*109 Bq (70 mCi). Herein we report the development of an electrochemical generator for the supply of 213Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1 M hydrochloric acid at 70 °C. Using this method, we were able to plate an average of 73 ± 4 % of the 213Bi in solution and obtain a final 213Bi recovery of 65 ± 8 % in 0.1 M citrate pH 4.5 via reverse electrolysis using titanium as the cathode. The recovered 213Bi had an average radiochemical purity of >99.8 % and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1 % (not optimized).

A combined inorganic-organic titanium-44/scandium-44g radiochemical generator.

Scandium-44g (t1/2 = 4.0 h) is an emerging radioisotope for positron emission tomography. It can be produced with a radiochemical generator using its long-lived parent, titanium-44 (t1/2 = 59.1 years). This work presents a new inorganic substrate for 44Ti/44gSc radiochemical generator design based on porous TiO2 microbeads (80 µm and 110 µm particle size, 60 Å pores). Comprehensive evaluation of conditions optimal for generator construction (44Ti loading) and use (44gSc elution) is provided in three steps. For stable 44Ti loading onto titania, heat-treatment at 180 °C for 90 min is shown to be effective while 0.3 M HCl(aq) is identified as the medium of choice for 44gSc elution. Two titania-based 3.6 MBq generators prepared under optimized conditions are characterized with respect to 44gSc recovery and 44Ti breakthrough. Each of these generators employed a different guard substrate to minimize 44Ti breakthrough, TiO2 microbeads and ZR resin. Both are shown to provide comparable 44gSc recoveries close to 50% but differ in 44Ti breakthrough, which is significantly lower with the organic ZR resin guard substrate at 0.0002%. This concept represents a new inorganic-organic approach to 44Ti/44gSc generator design. Benefits of both substrates are exploited: TiO2 has potential for durability necessary for utilizing the long half-life of the 44Ti parent while ZR resin guard segments minimize 44Ti breakthrough.

Engineering a modular 44Ti/44Sc generator: eluate evaluation in preclinical models and estimation of human radiation dosimetry.

BACKGROUND: 44Sc/47Sc is an attractive theranostic pair for targeted in vivo positron emission tomographic (PET) imaging and beta-particle treatment of cancer. The 44Ti/44Sc generator allows daily onsite production of this diagnostic isotope, which may provide an attractive alternative for PET facilities that lack in-house irradiation capabilities. Early animal and patient studies have demonstrated the utility of 44Sc. In our current study, we built and evaluated a novel clinical-scale 44Ti/44Sc generator, explored the pharmacokinetic profiles of 44ScCl3, [44Sc]-citrate and [44Sc]-NODAGA (1,4,7-triazacyclononane,1-glutaric acid-4,7-acetic acid) in naïve mice, and estimated the radiation burden of 44ScCl3 in humans. METHODS: 44Ti/44Sc (101.2 MBq) in 6 M HCl solution was utilized to assemble a modular ZR resin containing generator. After assembly, 44Sc was eluted with 0.05 M HCl for further PET imaging and biodistribution studies in female Swiss Webster mice. Based on the biodistribution data, absorbed doses of 44/47ScCl3 in human adults were calculated for 18 organs and tissues using the IDAC-Dose software. RESULTS: 44Ti in 6 M HCl was loaded onto the organic resin generator with a yield of 99.97%. After loading and initial stabilization, 44ScCl3 was eluted with 0.05 M HCl in typical yields of 82.9 ± 5.3% (N = 16), which was normalized to the estimated generator capacity. Estimated generator capacity was computed based on elution time interval and the total amount of 44Ti loaded on the generator. Run in forward and reverse directions, the 44Sc/44Ti ratio from a primary column was significantly improved from 1038 ± 440 to 3557 ± 680 (Bq/Bq) when a secondary, replaceable, ZR resin cartridge was employed at the flow outlet. In vivo imaging and ex vivo distribution studies of the reversible modular generator for 44ScCl3, [44Sc]-citrate and [44Sc]-NODAGA show that free 44Sc remained in the circulation significantly longer than the chelated 44Sc. The dose estimation of 44ScCl3 reveals that the radiation burden is 0.146 mSv/MBq for a 70 kg adult male and 0.179 mSv/MBq for a 57 kg adult female. Liver, spleen and heart wall will receive the highest absorbed dose: 0.524, 0.502, and 0.303 mGy/MBq, respectively, for the adult male. CONCLUSIONS: A clinical-scale 44Ti/44Sc generator system with a modular design was developed to supply 44ScCl3 in 0.05 M HCl, which is suitable for further radiolabeling and in vivo use. Our data demonstrated that free 44ScCl3 remained in the circulation for extended periods, which resulted in approximately 10 times greater radiation burden than stably chelated 44Sc. Stable 44Sc/47Sc-complexation will be more favorable for in vivo use and for clinical utility.

Production of zirconium-88 via proton irradiation of metallic yttrium and preparation of target for neutron transmission measurements at DICER.

A process for the production of tens to hundreds of GBq amounts of zirconium-88 (88Zr) using proton beams on yttrium was developed. For this purpose, yttrium metal targets (≈20 g) were irradiated in a ~16 to 34 MeV proton beam at a beam current of 100-200 µA at the Los Alamos Isotope Production Facility (IPF). The 88Zr radionuclide was produced and separated from the yttrium targets using hydroxamate resin with an elution yield of 94(5)% (1σ). Liquid DCl solution in D2O was selected as a suitable 88Zr sample matrix due to the high neutron transmission of deuterium compared to hydrogen and an even distribution of 88Zr in the sample matrix. The separated 88Zr was dissolved in DCl and 8 µL of the obtained solution was transferred to a tungsten sample can with a 1.2 mm diameter hole using a syringe and automated filling station inside a hot cell. Neutron transmission of the obtained 88Zr sample was measured at the Device for Indirect Capture Experiments on Radionuclides (DICER).

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.

Thorium chelators for targeted alpha therapy: Rapid chelation of thorium-226.

One of the main challenges in targeted alpha therapy is assuring delivery of the α-particle dose to the targeted cells. Thus, it is critical to identify ligands for α-emitting radiometals that will form complexes that are very stable, both in vitro and in vivo. In this investigation, thorium-227 (t1/2 = 18.70 days) chelation of ligands containing hydroxypyridinonate (HOPO) or picolinic acid (pa) moieties and the stability of the resultant complexes were studied. Chelation reactions were followed by reversed-phased HPLC and gamma spectroscopy. Studies revealed that high 227 Th chelation yields could be obtained within 2.5 h or less with ligands containing four Me-3,2-HOPO moieties, 1 (83%) and 2 (65%), and also with ligands containing pa moieties, H4 octapa 3 (65%) and H4 py4pa 6 (87%). No reaction occurred with H4 neunpa-p-Bn-NO2 4, and the chelation reaction with another pa ligand H4 pypa 5 gave inconsistent yields with a very broad radio-HPLC peak. The ligands spermine-(Me-3,2-HOPO)4 1, H4 octapa 3, and H4 py4pa 6 had high stability (i.e., 87% of 227 Th still bound to the ligand) in phosphate-buffered saline at room temperature over a 6-day period. Preliminary studies with ligand 6 demonstrated efficient chelation of thorium-226 (t1/2 = 30.57 min) when heated to 80°C for 5 min.

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.

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.

Alpha-Emitters and Targeted Alpha Therapy in Oncology: from Basic Science to Clinical Investigations.

Alpha-emitters are radionuclides that decay through the emission of high linear energy transfer α-particles and possess favorable pharmacologic profiles for cancer treatment. When coupled with monoclonal antibodies, peptides, small molecules, or nanoparticles, the excellent cytotoxic capability of α-particle emissions has generated a strong interest in exploring targeted α-therapy in the pre-clinical setting and more recently in clinical trials in oncology. Multiple obstacles have been overcome by researchers and clinicians to accelerate the development of targeted α-therapies, especially with the recent improvement in isotope production and purification, but also with the development of innovative strategies for optimized targeting. Numerous studies have demonstrated the in vitro and in vivo efficacy of the targeted α-therapy. Radium-223 (223Ra) dichloride (Xofigo®) is the first α-emitter to have received FDA approval for the treatment of prostate cancer with metastatic bone lesions. There is a significant increase in the number of clinical trials in oncology using several radionuclides such as Actinium-225 (225Ac), Bismuth-213 (213Bi), Lead-212 (212Pb), Astatine (211At) or Radium-223 (223Ra) assessing their safety and preliminary activity. This review will cover their therapeutic application as well as summarize the investigations that provide the foundation for further clinical development.

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.

Radiometric evaluation of diglycolamide resins for the chromatographic separation of actinium from fission product lanthanides.

Actinium-225 is a potential Targeted Alpha Therapy (TAT) isotope. It can be generated with high energy (≥ 100MeV) proton irradiation of thorium targets. The main challenge in the chemical recovery of 225Ac lies in the separation from thorium and many fission by-products most importantly radiolanthanides. We recently developed a separation strategy based on a combination of cation exchange and extraction chromatography to isolate and purify 225Ac. In this study, actinium and lanthanide equilibrium distribution coefficients and column elution behavior for both TODGA (N,N,N',N'-tetra-n-octyldiglycolamide) and TEHDGA (N,N,N',N'-tetrakis-2-ethylhexyldiglycolamide) were determined. Density functional theory (DFT) calculations were performed and were in agreement with experimental observations providing the foundation for understanding of the selectivity for Ac and lanthanides on different DGA (diglycolamide) based resins. The results of Gibbs energy (ΔGaq) calculations confirm significantly higher selectivity of DGA based resins for LnIII over AcIII in the presence of nitrate. DFT calculations and experimental results reveal that Ac chemistry cannot be predicted from lanthanide behavior under comparable circumstances.

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.