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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Production and performance of a 172 Hf/ 172 Lu generator

A 172 Hf/ 172 Lu radionuclide generator system to produce 172 Lu for laboratory scale applications in lutetium-based radiochemistry development was established and evaluated. The parent 172 Hf radionuclide was produced through 35.2 MeV proton irradiation of natural lutetium metal foil at the Brookhaven Linac Isotope Producer. Four resins were investigated for 172 Hf separation from bulk Lu target material: LN resin, ZR resin, in-house synthesized hydroxamate, and methyl-substituted hydroxamate resins, all with comparable performance. Separated 172 Hf was consolidated and used to create a ZR resin-based 4.9(3) MBq 172 Hf/ 172 Lu generator which was eluted 49 times over two years with no observed breakthrough of 172 Hf, and an average elution efficiency of 98(1)%. The eluted 172 Lu was used to radiolabel the macrocyclic chelator DOTA with an apparent molar activity of 8(2)x10 2 kBq/nmol.

172Hf/172Lu radionuclide generator↗
DOE OSTI2023

Evaluation of two extraction chromatography resins for scandium and titanium separation for medical isotope production

Scandium-47 ( 47 Sc) can be used in nuclear medicine as a therapeutic-diagnostic, or "theragnostic," radioactive medical isotope for cancer detection and treatment. The 47 Sc isotope can be produced through the photonuclear reaction 48 Ti(γ,p) 47 Sc by irradiating enriched 48 Ti target material. The enriched target material necessary for production is costly; 48 TiO 2 costs ~ $\$$1550/g, and targets can be > 50 g ($\$$77,500) to produce medically relevant amounts of 47 Sc. In order to keep costs low, a highly efficient separation of scandium from bulk titanium is desired, along with efficient methods for recycling the target material. Here this research is focused on evaluating efficient methods for the separation of scandium from bulk quantities of titanium using commercially available diglycolamide-based and hydroxamate-based extraction chromatography resins (DGA resin and ZR resin, respectively). The sorption of 47 Sc and Ti on these resins were investigated at varying concentrations of HNO 3 , HCl, H 2 SO 4 , and HF to explore how they might be used in a large-scale production/processing setting.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Improved titanium-44 purification process for establishing a high apparent molar activity titanium-44/scandium-44 generator

44 Sc-radiopharmaceuticals are gaining more interest but still lack availability. The proof of principle of a 44 Ti/ 44 Sc generator, which can produce 44 Sc daily, has been established but with some limitations and drawbacks. Despite recent advances, separation of 44 Ti from massive quantities of scandium target material is still cumbersome. Here, in this work, the improved radiochemical separation of 44 Ti from residual scandium target material was carried out by precipitation of Sc with fluoride ions. Furthermore, two approaches were used to set up a high apparent molar activity small-scale generator. The first method relied on extraction chromatography for fine purification using a DGA resin, followed by loading of the purified 44 Ti onto a ZR resin column. In the second method, 44 Ti was loaded on the ZR resin directly after the precipitation step. This second method was used to set up a generator of 370 kBq and evaluate by radiolabeling. An apparent molar activity of 2 MBq/nmol was obtained for the radiolabeling with DOTA, the most common and suitable chelate for scandium. This result is comparable with previously published data on 44m/44 Sc.

44Sc↗
DOE OSTI2023

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

Background: 44 Sc/ 47 Sc is an attractive theranostic pair for targeted in vivo positron emission tomographic (PET) imaging and beta-particle treatment of cancer. The 44 Ti/ 44 Sc 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 44 Sc. In our current study, we built and evaluated a novel clinical-scale 44 Ti/ 44 Sc generator, explored the pharmacokinetic profiles of 44 ScCl 3 , [ 44 Sc]-citrate and [ 44 Sc]-NODAGA (1,4,7-triazacyclononane,1-glutaric acid-4,7-acetic acid) in naïve mice, and estimated the radiation burden of 44 ScCl 3 in humans. Methods: 44 Ti/ 44 Sc (101.2 MBq) in 6 M HCl solution was utilized to assemble a modular ZR resin containing generator. After assembly, 44 Sc 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 / 47 ScCl 3 in human adults were calculated for 18 organs and tissues using the IDAC-Dose software. Results: 44 Ti in 6 M HCl was loaded onto the organic resin generator with a yield of 99.97%. After loading and initial stabilization, 44 ScCl 3 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 44 Ti loaded on the generator. Run in forward and reverse directions, the 44 Sc/ 44 Ti 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 44 ScCl 3 , [ 44 Sc]-citrate and [ 44 Sc]-NODAGA show that free 44 Sc remained in the circulation significantly longer than the chelated 44 Sc. The dose estimation of 44 ScCl 3 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 44 Ti/ 44 Sc generator system with a modular design was developed to supply 44 ScCl 3 in 0.05 M HCl, which is suitable for further radiolabeling and in vivo use. Our data demonstrated that free 44 ScCl 3 remained in the circulation for extended periods, which resulted in approximately 10 times greater radiation burden than stably chelated 44 Sc. Stable 44 Sc/ 47 Sc-complexation will be more favorable for in vivo use and for clinical utility.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2022

Anion exchange and extraction chromatography tandem column isolation of zirconium-89 ( 89 Zr) from cyclotron bombarded targets using an automated fluidic platform

The long-lived positron emitter 89 Zr is a highly promising nuclide employed in diagnostic Positron Emission Tomography (PET) imaging. Methods of radiochemical processing to obtain 89 Zr for clinical use are traditionally performed with a single hydroxamate resin column. In this work, we present a tandem column purification method for the preparation of high-purity 89 Zr from cyclotron bombarded Y metal foils. The primary column is a macroporous, strongly basic anion exchange resin on styrene divinylbenzene co-polymer. The secondary microcolumn, with an internal volume of 33 µL, is packed with an extraction chromatography resin (ExCR) loaded with HDEHP. A condition of “inverted selectivity” is presented, wherein the 89 Zr elution from the primary column is synonymous with the load condition on the secondary column. The ability to transfer 89 Zr from one column to the next allows two sequential column clean-up methods to be performed prior to the final elution of the 89 Zr product. This approach assures delivery of very high purity 89 Zr . The tandem column purification process has been implemented into a prototype automated fluidic system. Optimization of the method is presented, followed by evaluation of the process using seven cyclotron bombarded Y metal foil targets. Once optimized, we found that 93.7±2.3% of the 89 Zr present in the foils was recovered in the secondary column elution fraction (0.8 M oxalic acid). Radiochromatograms of the product elution peaks enabled determination of full width at half-maximum (FWHM) and 89 Zr collection yields as a function of volume. Because of the small size of the secondary microcolumn, a 89 Zr product volume of ~0.28 mL is reported, which provides a substantially increased nuclide concentration over traditional methods. Finally, we evaluated the transchelation of the resulting 89 Zr oxalate product to deferoxamine mesylate (DFOM) salt. We observed effective specific activities (ESA) and bindable metals concentrations ([M B ]) that exceed those reported by the traditional single hydroxamate column method.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Harvesting 88 Zr from heavy-ion beam irradiated tungsten at the National Superconducting Cyclotron Laboratory

Tungsten is a commonly used material at many heavy-ion beam facilities, and it often becomes activated due to interactions with a beam. Many of the activation products are useful in basic and applied sciences if they can be recovered efficiently. In order to develop the radiochemistry for harvesting group (IV) elements from irradiated tungsten, a heavy-ion beam containing 88 Zr was embedded into a stack of tungsten foils at the National Superconducting Cyclotron Laboratory and a separation methodology was devised to recover the 88 Zr. The foils were dissolved in 30% hydrogen peroxide, and the 88 Zr was chemically purified from the tungsten matrix and from other co-implanted radionuclides (such as 85 Sr and 88 Y) using strong cation-exchange (AG MP-50) chromatographic resin in sulfuric acid media. The procedure provided 88 Zr in approximately 60 mL 0.5 M sulfuric acid with no detectable radio-impurities. The overall recovery yield for 88 Zr was (92.3 ± 1.2)%. In conclusion, this proof-of-concept experiment has facilitated the development of methodologies to harvest from tungsten and tungsten-alloy parts that are regularly irradiated at heavy-ion beam facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

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

Abstract A process for the production of tens to hundreds of GBq amounts of zirconium-88 ( 88 Zr) 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 88 Zr radionuclide was produced and separated from the yttrium targets using hydroxamate resin with an elution yield of 94(5)% (1σ). Liquid DCl solution in D 2 O was selected as a suitable 88 Zr sample matrix due to the high neutron transmission of deuterium compared to hydrogen and an even distribution of 88 Zr in the sample matrix. The separated 88 Zr 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 88 Zr sample was measured at the Device for Indirect Capture Experiments on Radionuclides (DICER).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Large-scale production of 88 Y and 88 Zr/ 88 Y generators: A proof of concept study for a 70 MeV H – cyclotron

Here, the large-scale production of 88 Y with proton-induced reactions has been investigated from the perspective of new generation 70 MeV H – cyclotrons. Tandem target configurations are presented for both the direct production of 88 Y as well as for producing 88 Zr/ 88 Y generators. Based on the relevant excitation functions, physical yields have been derived for 88 Y production with Y 2 O 3 /SrCO 3 tandem targets and 88 Zr production with Zr/Y 2 O 3 tandem targets. Yields are presented for optimized targets (i.e. optimum yield) as well as for balanced thermal loads on the individual targets. Liquid 88 Zr/ 88 Y generators have been produced using both natural Zr and Nb target materials, the former for dedicated productions and the latter as a byproduct by processing spent irradiated Nb capsules which normally would constitute radioactive waste. These stock solutions, which contain both the target material and 88 Zr precursor, are retained virtually unchanged after processing except for the removal of 88 Y on AG MP-50 macroporous cation-exchange resin. Methods are presented for the preparation of Nb stock solutions in hydrofluoric acid and Zr stock solutions in sulfuric acid. It is shown that multi-Ci productions of 88 Y are feasible at a 70 MeV cyclotron facility, suitable for the needs of fracking applications. In addition, 88 Zr/ 88 Y generators can provide 88 Y with very high specific activity, suitable for labelling of biomolecules.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Digital light processing of porous LLZTO scaffolds for Li-garnet solid-state batteries

Li 7 La 3 Zr 2 O 12 (LLZO)-based solid-state electrolytes (SEs) are promising materials for next-generation solid-state batteries. In this work, digital light processing (DLP), an emerging additive manufacturing technology, is employed to produce porous Ta-doped LLZO (LLZTO) scaffolds. The self-standing scaffolds are 100 μm thick and have 40% porosity. The scaffolds demonstrate symmetric cell cycling stability exceeding 1,500 h at 0.1 mA/cm2 current density, with a capacity of 0.1 mAh/cm 2 (1 h for each half cycle). At higher current densities, reversible soft shorts frequently happen, while immediate hard shorts are prevented due to Li dendrite growth being hindered by the tortuous pore network. In addition to the cycling stability, the phase stability of LLZTO is investigated during the post-printing thermal process for printing resin removal. We discovered that the LLZTO partially decomposes into Li 2 Zr 2 O 7 and other impurity phases from 400°C to 800°C, but the pure LLZTO phase is restored upon the completion of resin removal beyond 800°C.

Li-metal anode↗
DOE OSTI2020

Scaling Trivalent Actinide and Lanthanide Recovery by Diglycolamide Resin from Savannah River Site’s Mark-18A Targets

The Savannah River National Laboratory is implementing a separation flowsheet to recover rare isotopes and transplutonium elements from irradiated 242 Pu targets known as Mark-18A targets. The Mark-18A targets contain the United States’ supply of nonseparated 244 Pu, which has a wide range of applications from nuclear forensics to production of superheavy elements such as flerovium. The targets also contain hundreds of grams of heavy curium ( 246–248 Cm), which is used as a target material for 252 Cf production. This work investigates the use of diglycolomide resin (DGA Resin) to recover valuable trivalent actinides from the Mark-18A targets. Batch contact experiments were performed on a representative simulant to determine mass loadings. Furthermore, the resin showed an overall capacity of 11 mg/mL for a mixed metal matrix (Zr and La–Gd). Column experiments showed chromatographic separation with transition-metal breakthroughs occurring first followed by the lanthanide series La–Gd. The experiments showed that lanthanide breakthrough occurred after 11 mg/mL mass loading was reached on the column with the mixed metal matrix. A radiological column experiment with an in-line UV/vis cell was able to detect Nd breakthrough just prior to 241 Am breakthrough. Implementing an in-line UV/vis cell into full-scale Mark-18A target processing will be used to limit breakthrough of trivalent actinides recovered from the targets.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Methods for separating isotopes from a sample of fission products

Systems and methods for efficient, effective, and safe separation and isolation of multiple isotopes (e.g., Mo, Zr, Ba, Sr, Te, and lanthanide isotopes) from fission products includes use of a plurality of chromatography columns, each containing a chromatographic resin formulated to target one or more particular isotopes. The system is operable in a “series” configuration to load the multiple columns by a single pass of the sample. Then, the system may be transitioned (e.g., using valves) to a “parallel” configuration in which multiple columns of the system may be operated simultaneously to elute targeted isotopes. Additional parallel operations of the columns, using different eluent compositions, may be used to elute different targeted isotopes. The system may be reconditioned in preparation for a subsequent sample.

Snow, Mathew S.↗
DOE OSTI2023

Systems for separating isotopes from a sample of fission products

Systems and methods for efficient, effective, and safe separation and isolation of multiple isotopes (e.g., Mo, Zr, Ba, Sr, Te, and lanthanide isotopes) from fission products includes use of a plurality of chromatography columns, each containing a chromatographic resin formulated to target one or more particular isotopes. The system is operable in a “series” configuration to load the multiple columns by a single pass of the sample. Then, the system may be transitioned (e.g., using valves) to a “parallel” configuration in which multiple columns of the system may be operated simultaneously to elute targeted isotopes. Additional parallel operations of the columns, using different eluent compositions, may be used to elute different targeted isotopes. The system may be reconditioned in preparation for a subsequent sample.

Snow, Mathew S.↗
DOE OSTI2022

Development of Chemical Procedures for Isotope Harvesting: Separation of Trace Hafnium from Tungsten

The separation of trace hafnium from bulk tungsten alloys is of interest for isotope harvesting at the National Superconducting Cyclotron Laboratory and the Facility for Rare Ion Beams because 172 Hf, the parent of 172 Lu, accumulates in tungsten alloy beam blockers at these facilities. In this work, a procedure for the separation of trace hafnium from a bulk tungsten alloy (454 g) was established using tracer isotopes ( 175 Hf, 88 Zr, 173 Lu, and 88 Y). Here, the procedure employed dissolution in an HF–HNO 3 solution followed by a calcium fluoride precipitation, and then extraction chromatography was used for more selective separation steps. Two stages of column separations using LN resin (HDEHP based) and TRU resin (CMPO based) were performed. Gamma-ray spectroscopy and mass spectrometry were used to analyze the final hafnium sample and follow the steps in the chemical processing. Furthermore, the final recovery of hafnium was high (90 ± 8)%, and the mass of tungsten and other transition metals was reduced to near background levels (as determined by ICP-MS of the blank acid solutions). Zirconium follows hafnium quantitatively in this procedure; there was no detectable 173 Lu in the final hafnium sample.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗