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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Ra-226 extraction and purification from simulated uranium tailings

Growing demand for Ac-225 has outstripped current production, increasing interest in alternative production routes. Furthermore, these often rely on irradiations of Ra-226, yet Ra-226 supplies are also limited. Uranium tailings provide an abundant source of Ra-226, though separation from these complex matrices remains challenging. A laboratory-scale process was developed to recover Ra-226 from simulated U-tailings. After acid leaching, Ra remained adsorbed on refractory oxides and silicates. An additional CaCl₂ leach of the residue raised recovery to ~ 64%. Subsequent purifications, including precipitation, ion-exchange, and extraction chromatography (TK101), achieved chemical yields up to 73 ± 7%, producing 7.68 ± 0.1 µCi of high-purity Ra-226.

and nuclear chemistry↗
DOE OSTI2025

Separation of terbium from proton-irradiated gadolinium oxide targets – development of an effective, scalable and automatable process

This work reports an effective and scalable radiochemical separation process for isolating terbium from Gd 2 O 3 . The separation process uses three commercially available extraction chromatography resin columns, has been implemented on a computer-controlled chemistry module, and tested with 100 mg quantities of proton-irradiated nat Gd 2 O 3 . Here, the 4-hour separation procedure isolated radioterbium in 1.3 mL of 0.01 M HCl with 80 ±8% radiochemical yield and a Gd decontamination factor >(1.2 ± 0.3)·10 5 .

Adjacent lanthanide separation↗
DOE OSTI2021

A High Separation Factor for 165Er from Ho for Targeted Radionuclide Therapy

Background: Radionuclides emitting Auger electrons (AEs) with low (0.02–50 keV) energy, short (0.0007–40 µm) range, and high (1–10 keV/µm) linear energy transfer may have an important role in the targeted radionuclide therapy of metastatic and disseminated disease. Erbium-165 is a pure AE-emitting radionuclide that is chemically matched to clinical therapeutic radionuclide 177Lu, making it a useful tool for fundamental studies on the biological effects of AEs. This work develops new biomedical cyclotron irradiation and radiochemical isolation methods to produce 165Er suitable for targeted radionuclide therapeutic studies and characterizes a new such agent targeting prostate-specific membrane antigen. Methods: Biomedical cyclotrons proton-irradiated spot-welded Ho(m) targets to produce 165Er, which was isolated via cation exchange chromatography (AG 50W-X8, 200–400 mesh, 20 mL) using alpha-hydroxyisobutyrate (70 mM, pH 4.7) followed by LN2 (20–50 µm, 1.3 mL) and bDGA (50–100 µm, 0.2 mL) extraction chromatography. The purified 165Er was radiolabeled with standard radiometal chelators and used to produce and characterize a new AE-emitting radiopharmaceutical, [165Er]PSMA-617. Results: Irradiation of 80–180 mg natHo targets with 40 µA of 11–12.5 MeV protons produced 165Er at 20–30 MBq·µA−1·h−1. The 4.9 ± 0.7 h radiochemical isolation yielded 165Er in 0.01 M HCl (400 µL) with decay-corrected (DC) yield of 64 ± 2% and a Ho/165Er separation factor of (2.8 ± 1.1) · 105. Radiolabeling experiments synthesized [165Er]PSMA-617 at DC molar activities of 37–130 GBq·µmol−1. Conclusions: A 2 h biomedical cyclotron irradiation and 5 h radiochemical separation produced GBq-scale 165Er suitable for producing radiopharmaceuticals at molar activities satisfactory for investigations of targeted radionuclide therapeutics. This will enable fundamental radiation biology experiments of pure AE-emitting therapeutic radiopharmaceuticals such as [165Er]PSMA-617, which will be used to understand the impact of AEs in PSMA-targeted radionuclide therapy of prostate cancer.

165Er↗
DOE OSTI2026

Transition Metals Separation with Commercial Neutral Extractants – A Review

The increasing use of extraction chromatography resins across fields such as hydrometallurgy, nuclear medicine, and environmental analysis has created a need for a deeper understanding of their interactions with transition metals. Despite extensive research on f-element separations, the behavior of transition metals in these systems remains relatively understudied. This review provides a comprehensive overview of the current state of knowledge on the extraction behavior of transition metals with neutral extractants, including TODGA, TEHDGA, TBP, and CMPO, and their corresponding resins, such as DGA, BDGA, UTEVA, TBP, and TRU. The review summarizes extraction data, extracted complex coordination environments, separation reaction stoichiometries, and associated thermodynamics, highlighting inconsistencies and knowledge gaps in the literature. The study emphasizes the need for further research using spectroscopy and computational methods to elucidate extraction mechanisms and to improve the efficiency and selectivity of transition metal separations. By identifying areas for future research and development, this review aims to stimulate advancements in the field and promote the development of innovative separation technologies. The implications of this research are far-reaching, with potential applications in nuclear waste management, nuclear forensics, metal recovery, and environmental remediation. Overall, this review provides a foundation for future studies on the extraction of transition metals using neutral extractants and resins.

Wall, Nathalie A.↗
DOE OSTI2025

Online collection of LC effluent as droplets for automatic injection into ICP-MS

A sample introduction technique was developed to allow for online, dropwise injection of effluent into a mass spectrometer (MS). This allowed for the coupling of chromatography systems that were not driven by pumps without altering the set-up or separation performance. This capability combined the benefits of coupling separation and measurement while also enabling accurate chromatographic evaluation that would generally be performed offline. The incorporation of a flowing rinse into the system made dropwise resolution possible for any liquid capable of forming drops under ambient conditions. The technique was demonstrated utilizing gas pressurized extraction chromatography (GPEC) and inductively coupled plasma-time-of-flight mass spectrometry (ICP-TOFMS) to illustrate the qualitative applications for rapid separation development and procedure evaluation. This method's quantitative applications were evaluated using single and double isotope dilution mass spectrometry (IDMS) with an external mass bias correction to measure analytes across entire elutions including volumes as small as single drops. In conclusion, the total neodymium values afforded by both single and double IDMS were within the uncertainty of the calculated value for evaluation of full elution peaks and for most of the single drops analyzed that reached above 10k cps Nd-144.

GPEC↗
DOE OSTI2025

Rapid separation of radiopure yttrium-91 for tracer studies from mixed fission products and uranium

A novel, single step method for isolating 91 Y from irradiated uranium and mixed fission products has been developed based on the commercially available Eichrom LN resin. The separation procedure allows for loading an LN resin column with an irradiated uranium solution, containing mixed fission products, in dilute HCl where both uranium and yttrium are retained on the resin. Eluting dilute (0.5–1 M) HCl will strip the majority of fission products. Furthermore, increasing the concentration to 3 M HCl will elute 91 Y without the presence of any other fission products. Finally, uranium can be recovered by passing concentrated HCl through the column.

Chromatography↗
DOE OSTI2025

Photonuclear production of 225 Ac from 226 Ra targets

Herein we report proof of principal production of 225 Ac via 226 Ra(γ , n) 225 Ra $\overset{^-β}{\rightarrow}$ 225 Ac and subsequent chemical separation via extraction chromatography using BDGA and Ln resins. Irradiation of 0.35 µg 226 Ra with 39 MeV bremsstrahlung end-point energy photons produced 225 Ra and 224 Ra at a rate of 754 and 2215 kBq/g 226 Ra/mA/h, respectively. The production rate of 225 Ac determined at its peak radioactivity was 335 kBq/g 226 Ra/mA/h. Radiochemically pure 225 Ac was isolated with a minimum separation factor of 125,000 for 225 Ac: 226 Ra. In conclusion, these results are promising and lay the foundation for future studies to scale-up production for preclinical quantities of high purity 225 Ac using the 226 Ra(γ , n) 225 Ra $\overset{^-β}{\rightarrow}$ 225 Ac approach.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
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 OSTI2025

48 Ca(p,pn) 47 Ca linac production for the preparation of a 47 Ca/ 47 Sc generator

Here, the production of 47 Sc via the nuclear reaction 48 Ca(p,pn) 47 Ca→(β-, T 1/2 = 4.54 d) 47 Sc and the assembly of a 47 Ca/ 47 Sc generator system has been investigated at Brookhaven National Laboratory. To produce 47 Ca, a pressed nat CaCl 2 target was irradiated with a proton beam at energy on target of 27.5 MeV at Brookhaven Linac Isotope Producer. Irradiated targets were dissolved in water and then acidified with concentrated HCl before separation using extraction chromatography through a diglycolamide (DGA) resin column. After ingrowth, 47 Sc was collected at >90 % with low contamination from both nonradioactive metals and coproduced radionuclides. A total of four 47 Ca/ 47 Sc separations were performed with 6–7 days between the irradiation and the separation to allow the equilibration of the ingrowing 47 Sc.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Processing iridium foil targets for the production of radioplatinum isotopes

Dissolving noble metals remains challenging because of their high reduction potentials and tendency to passivate as oxides. This barrier is acute for iridium metal targets used to produce platinum radioisotopes, where post-irradiation processing is costly and hazardous with conventional routes. Here, we report a rapid alkali-fusion method that converts Ir metal directly into chloride media compatible with downstream separations. In the process, an Ir foil is fused with Na 2 O 2 and KCl at temperatures above the KCl melting point; the molten salt enhances heat transfer and facilitates O 2− /Na + transport to the Ir surface, overcoming Ir's oxidation resistance. The fused cake is then cooled and leached in 10 M HCl, yielding Ir in solution without specialized microwave digestion or HF. Pt separation from bulk Ir was evaluated using extraction-chromatography resins. A newly formulated ionic-liquid resin containing Cyphos IL 104 (E104) was compared with two commercial resins (TrisKem TK200 and TK201). All resins achieved >90 % Ir recovery. Pt co-recovery depended on resin chemistry: TK200 recovered 85.9 % Pt, TK201 recovered 54.1 % Pt, and E104 recovered 72.9 % Pt. This study introduces a fast, scalable dissolution route for solid Ir targets and demonstrates resin choices that tune Ir/Pt separation, lowering the barrier to routine production and purification of Pt radioisotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Compact automated apparatus for rapid astatine recovery from nitric acid media: Design, application, and impurity characterization

We report automation of irradiated bismuth target dissolution and astatine recovery from nitric acid media has been achieved at Texas A&M University. This process can be controlled remotely; it does not require any chemical treatment or evaporation of dissolution solution prior to the separation of above-mentioned elements on an extraction chromatography column, making the final product ready within 20 min. The accompanying radio-impurities have been identified and successfully separated from astatine, leading to greater than 99% radiopurity of the desired fractions. The system consists of only one pump and a set of LabVIEW controlled valves, and allows the connection of up to 10 different columns, providing the opportunity to prepare a variety of air dry columns, which can be different column geometries, containing astatine ready for further chemistry or shipment to desired facilities.

42 ENGINEERING↗
DOE OSTI2020

Innovative high-temperature ammonium bifluoride fusion and rapid analysis of elements with nuclear forensic value

High-temperature ammonium bifluoride (ABF) fusions were evaluated for potential use in rapid dissolution of post-detonation nuclear debris. The ABF fusion was carried out in a Pt crucible which allowed evaluation of higher fusion and evaporation temperatures. The high-temperature ABF fusion dissolution method was evaluated using geological reference materials: USGS QLO-1a Quartz Latite, USGS SDC-1 Mica Schist, and NIST 278 Obsidian Rock. The optimized dissolution method involved a 10 min fusion at 540 °C, a 5 min reflux in 8 M HNO 3 , an evaporation at 300 °C and final dilution into 45 mL of 2% (v/v) HNO3. The final solution was filtered after heating at 105 °C using a hotblock. This dissolution method was simple, requiring only a hotplate or hotblock, filtered samples were available for ICP-MS analysis or radiochemical separation within 150 min, and was found to have high (>90%) recovery for many isotopes of interest in nuclear forensics applications. U and Pu in the dissolved material was separated using TEVA and UTEVA extraction chromatography columns, a process which resulted in >90% recovery. An irradiated U tracer was spiked into the material prior to dissolution and analyzed for recovery of major fission products and 239Np. The monitored radionuclides had recoveries of greater than 90%, except for the volatile radioiodine isotopes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Multiscale investigations of europium(III) complexation with tetra- n -octyl diglycolamide confined in porous solid supports

The microscopic, short-range coordination environments and mesoscopic, long-range structures of Eu 3+ contacted with tetra-n-octyl diglycolamide (TODGA) confined on ordered mesoporous carbon (OMC) nanoparticles and Amberchrom CG 71 resin were probed using Eu L 3 -edge extended X ray absorption fine structure (EXAFS) as well as small-angle and wide-angle X ray scattering, SAXS and WAXS, respectively. A homoleptic Eu(TODGA) 3 3+ coordination complex typical of liquid-liquid extraction (LLE) chemistry is present under low Eu 3+ loading conditions on both solid supports. Deviations from this traditional structure motif appear at hyperstoichiometric Eu 3+ loadings, above the 1:3 Eu 3+ to TODGA mole ratio. Microcrystalline-like domains with multinuclear Eu speciation are templated by use of these high Eu 3+ loading conditions with the functionalized OMC materials, highlighting a major departure of liquid-solid extraction chromatography (EXC) from the coordination chemistry of LLE. No such long-range spatial coherence was observed for analogous polyacrylic resin materials. Furthermore, these results demonstrate both the similarities and differences between multiscale structures in LLE and EXC, underscoring the opportunities for improved separation techniques based on solid supports with long-range spatial coherence (e.g., OMC systems) and without it (e.g., resin materials). Since crystallization was the first approach to adjacent lanthanide separations, the prospect of templating microcrystalline TODGA-lanthanide complexes that are otherwise not prone to crystallization by use of OMC nanoparticles represents a new entry to addressing long-standing issues in purification by liquid-solid phase separation.

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

Measuring Signatures of Fuel Irradiation in Large Particle Samples

Understanding the operating history of a nuclear facility is one of the key goals in nuclear safeguards, enabling the verification of declared activities. This is accomplished via a number of mechanisms, including the analysis of samples collected at the facility. The best metrics for understanding and verifying the history of nuclear fuel are (1) actinide isotopic content and (2) actinide inter-element ratios. While techniques have long existed for measuring these values with high accuracy and precision in bulk samples, as samples become smaller – such as may be collected in environmental collection – new methods are required both for clean processing of pure samples, and for analysis of these small samples. This study describes a new extraction chromatography technique for recovering U, Pu, and Am from single small aliquots, and advanced mass spectrometry methods for high-precision and high-accuracy analysis of trace and ultra-trace levels of actinides. These techniques are applied to particle samples cut from fuel from a historic nuclear reactor. Ages calculated from Pu–Am chronometry of particles are in excellent agreement with the known operating history of the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Ion exchange behavior of astatine and bismuth

Astatine and bismuth sorption on several ion exchange resins from nitric acid media has been studied. This work covers commercially available resins Dowex 50 × 4, MP thiol, Dowex 1 × 8, and TEVA. One of the main advantages of using ion exchangers for astatine separation and purification is the absence of any organic media in the effluent in comparison with extraction chromatography resins. The behavior of the above-mentioned metals was investigated in up to 4 M HNO 3 solutions. The determined distribution coefficients are greater than 1 for all the resins studied, reaching 10 4 for MP thiol and TEVA resins. The interaction of nitric acid with an anion exchanger in Cl-form results in an exchange reaction, where AtO + chloro complexes can form. To understand the astatine behavior under these conditions, stability constants of AtOCl and AtOCl 2 – complexes have been reevaluated and discrepancy in the literature values has been eliminated. For each resin a thermodynamic model has been developed to suggest a possible mechanism of astatine sorption. Literature and new experimental data on bismuth sorption by studying resins in nitric acid media have been reviewed and a mathematical model to describe its behavior has been suggested. Furthermore, a ratio of corresponding fit functions of At and Bi assigned to the same resin and acidity has been used to estimate the separation factors of these elements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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↗
DOE OSTI2020

Recycling Isotopically Enriched Mo from the RadioGenix Waste Stream

NorthStar’s RadioGenix generator utilizes an aqueous biphasic extraction chromatography column to separate Tc from low-specific-activity 99 Mo. During this separation process, a fraction of the source Mo material can be found in the waste. This is not of concern when natural Mo targets are used. However, the introduction of isotopically enriched 98 Mo for neutron capture or 100 Mo for accelerator-based production requires recovery of this valuable enriched Mo material. The generator waste is absorbed and solidified in a superabsorbent hydrogel polymer. In this contribution, we report results from our small-scale experimental study on development of a separation process for recovery of molybdenum from the polymeric gel waste using common chemicals and simple chemical techniques. Our investigation confirms that the acidic deswelling of the alkaline hydrogel with concentrated hydrochloric acid converts the occluded molybdate to a neutral molybdenyl dichloride species that is freely released from the gel. Mo can then be concentrated on a commercial tri-n-butyl phosphate resin, eluted with ammonium hydroxide, and crystallized as ammonium heptamolybdate for conversion back to Mo metal. Typical Mo recoveries are 90-95%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Cyclotron production of 43Sc and 44gSc from enriched 42CaO, 43CaO, and 44CaO targets

Introduction: 43 Sc and 44g Sc are both positron-emitting radioisotopes of scandium with suitable half-lives and favorable positron energies for clinical positron emission tomography (PET) imaging. Irradiation of isotopically enriched calcium targets has higher cross sections compared to titanium targets and higher radionuclidic purity and cross sections than natural calcium targets for reaction routes possible on small cyclotrons capable of accelerating protons and deuterons. Methods: In this work, we investigate the following production routes via proton and deuteron bombardment on CaCO 3 and CaO target materials: 42 Ca(d,n) 43 Sc, 43 Ca(p,n) 43 Sc, 43 Ca(d,n) 44g Sc, 44 Ca(p,n) 44g Sc, and 44 Ca(p,2n) 43 Sc. Radiochemical isolation of the produced radioscandium was performed with extraction chromatography using branched DGA resin and apparent molar activity was measured with the chelator DOTA. The imaging performance of 43 Sc and 44g Sc was compared with 18 F, 68 Ga, and 64 Cu on two clinical PET/CT scanners. Discussion: The results of this work demonstrate that proton and deuteron bombardment of isotopically enriched CaO targets produce high yield and high radionuclidic purity 43 Sc and 44g Sc. Laboratory capabilities, circumstances, and budgets are likely to dictate which reaction route and radioisotope of scandium is chosen.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗