Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “NUCLEAR MEDICINE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Experiments with a prototype titanium hot cavity surface ionization source intended for electromagnetic separation of radioactive samarium and other lanthanide elements

This paper reports experimental results of a prototype titanium surface ionization source. For the first time, a lanthanide ion beam has been produced with a surface ionizer composed completely of titanium metal. Titanium does not readily activate with neutron irradiation. This offers the potential for inserting an ion source made of titanium directly into a reactor with a pre-loaded non-radioactive lanthanide target. This seamlessly integrates target irradiation with isotope separation, eliminating post irradiation sample manipulation. Samarium ion beam currents up to 960 nA have been produced in an off-line test bench equipped with rudimentary beam optics. This is a crucial step toward the development of an ionization source adopted for the electromagnetic isotope separator (EMIS) facility, which has been designed for high throughput separations of radioactive 153 Sm and other lanthanides of interest in the field of nuclear medicine. The ion current and important factors affecting the performance of the ion source, such as the ionizer temperature and thermal gradient, are discussed. Lastly, the experimental results are presented together with a discussion of future modifications to optimize the overall surface ionization source performance.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Nuclear Data for Spallation Neutron Radioisotope Production (Final Technical Report)

Over 50 million nuclear medicine procedures are performed annually, leading to a multi-billion dollar market for radioisotope production. The demand for new medical and research isotopes is growing, and radioisotope supply is insufficient. Most radioisotope production today utilizes charged particle or low-energy neutron irradiation of a target. Isotope production with tens to hundred MeV incident energies is a relatively unexplored option. There is a tremendous opportunity associated with a growing number of suitable domestic and international facilities buttressed by hundred million dollar global investments (e.g., the Los Alamos and Brookhaven Isotope Production Facilities, the European Spallation Source in Lund, and the Korean Multi-purpose Accelerator Complex in Gyeongbuk). In part due to a lack of supporting nuclear data that would make modeling radioisotope yields and purities possible, these facilities do not utilize their high-energy neutron fluxes for isotope production. This project attempted to measure neutron reaction excitation functions relevant to the large-scale production of critical radioisotopes, enabling development of cost-efficient isotope production methods, contributing to the improvement of theoretical models, and enhancing the value of national isotope production facilities. Reactions which form 67 Cu, 32 Si, and alpha-emitting isotopes like 225 Ac were chosen for their consistent prioritization by expert panels, representation of diverse reaction mechanisms, and relative lack of supporting nuclear data. Accurate measurement of these data is presently made only using quasi-monoenergetic neutron beams, which are produced by bombarding thin lithium targets with protons at only a few laboratories in the world. Ultimately, the projects efforts were incompletely accomplished due to a combination of multi-year technical issues at the only laboratory in the world which can make the proposed measurements, iThemba Labs in South Africa, and the onset of a global pandemic, which precluded the significant effort and time required to develop an independent measurement capability in the United States.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Separation and purification of Mo-99 produced from natural U 3 0 8 targets via photo-fission

The most-used medical isotope is 99m Tc (t½ = 6.0 h), which comprises over 80% of isotopes used in nuclear medicine today. It is normally derived from its transient equilibrium parent 99 Mo (t½ = 66 h). A recent surge of interest in using the more proliferation-resistant low-enriched uranium (LEU) under the American Medical Isotope Production Act (AMIPA) has presented this technical challenge: can a domestic supplier meet the estimated weekly U.S. demand of 1500 6–day Ci of 99 Mo?

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2022

Cutaneous and local radiation injuries

The threat of a large-scale radiological or nuclear (R/N) incident looms in the present-day climate, as noted most recently in an editorial in Scientific American (March 2021). These large-scale incidents are infrequent but affect large numbers of people. Smaller-scale R/N incidents occur more often, affecting smaller numbers of people. There is more awareness of acute radiation syndrome (ARS) in the medical community; however, ionising radiation-induced injuries to the skin are much less understood. Here, this article will provide an overview of radiation-induced injuries to the skin, deeper tissues, and organs. The history and nomenclature; types and causes of injuries; pathophysiology; evaluation and diagnosis; current medical management; and current research of the evaluation and management are presented. Cutaneous radiation injuries (CRI) or local radiation injuries (LRI) may lead to cutaneous radiation syndrome, a sub-syndrome of ARS. These injuries may occur from exposure to radioactive particles suspended in the environment (air, soil, water) after a nuclear detonation or an improvised nuclear detonation (IND), a nuclear power plant incident, or an encounter with a radioactive dispersal or exposure device. These incidents may also result in a radiation-combined injury; a chemical, thermal, or traumatic injury, with radiation exposure. Skin injuries from medical diagnostic and therapeutic imaging, medical misadministration of nuclear medicine or radiotherapy, occupational exposures (including research) to radioactive sources are more common but are not the focus of this manuscript. Diagnosis and evaluation of injuries are based on the scenario, clinical picture, and dosimetry, and may be assisted through advanced imaging techniques. Research-based multidisciplinary therapies, both in the laboratory and clinical trial environments, hold promise for future medical management. Great progress is being made in recognising the extent of injuries, understanding their pathophysiology, as well as diagnosis and management; however, research gaps still exist.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Production of Medically Desirable Radioisotopes in the EIRENE Molten Salt Reactor

Radioisotopes play a vital role in nuclear medicine, enabling the performance of diagnostic procedures such as Positron Emission Tomography (PET) and Single Photon Emission Computed Tomography (SPECT), in addition to innovative targeted therapies for selectively delivering cytotoxic radiation doses to tumor cells while minimizing damage to healthy tissue [1]. Beta emitting radionuclides such as 90Y and 131I find wide use in radio-immunotherapy, with the radiopharmaceuticals 90Y-ibritumomab tiuxetan (Zevalin®) and 131I-tositumomab (Bexxar®) having received FDA approval for treatment of non-Hodgkin’s lymphoma through targeting the CD20 surface receptor, which is commonly expressed in many B cell non-Hodgkin’s lymphoma subtypes [2, 3]. 90Y has also been successfully applied for treatment of brain tumors [4, 5] and liver cancer [6]. The favorable chelation chemistry of 90Y enables its use with a variety of ligands for the development of radiopharmaceuticals suitable for targeting different carcinoma types [1].

42 - ENGINEERING↗
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 OSTI2020

Current Status of the Synroc Waste Treatment Facility - 20263

Construction of the Synroc Waste Treatment Facility is well underway at ANSTO. This plant has been designed to treat intermediate level liquid waste arising from the production of Mo-99 within the newly constructed ANSTO Nuclear Medicine (ANM) precinct [1]. The facility utilises ANSTO Synroc technology that has been tailored to the chemical, physical, and radiological properties of the waste. The result is a highly durable wasteform with a significant reduction in the final volume of the treated waste. This paper presents the construction status of the Synroc Waste Treatment Facility and the status of the associated technology maturation plan. Construction of the building is scheduled for completion in late 2020 followed by process installation and commissioning. A key component of the technology maturation plan has been the construction of an Inactive Engineering Demonstrator [2]. This has aided in the mitigation of risks with respect to technology selection, process integration, and process nuclearisation. Furthermore, the demonstration facility has provided an environment for the design and development of the instrumentation and control philosophy resulting in a seamlessly integrated plant control system. Successful demonstration of the technology by the inactive engineering demonstrator has significantly reduced the associated risks. Details of the Synroc Waste Treatment Facility project progress will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Purification and Concentration of 99 Mo from a UREX raffinate

The medical isotope 99m Tc ( t ½ = 6.0 h) accounts for over 80% of isotopes used in diagnostic nuclear medicine today and is normally derived from its transient equilibrium parent 99 Mo ( t ½ = 66 h). To reduce U.S. dependence on 99 Mo derived from HEU by foreign suppliers, the NNSA/M3 program—under the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not utilize HEU. Superconducting electron linear accelerators with high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and LEU fission. After sufficient production intervals, targets can be rotated out and processed while another batch is irradiated. The process flow chemistry is shown in Figure 1. The irradiated U 3 O 8 targets are retrieved and dissolved in HNO 3 ; the volatile fission products are expelled and captured during this step. The HNO 3 liquor bearing the uranyl, fission and activation products is injected into the UREX suite of liquid-liquid extraction banks. The uranyl is partitioned in typical fashion using tri-$\textit{n}$-butyl phosphate (TBP) in a hydrocarbon diluent. The raffinate—comprising mainly 99 Mo and fission products—is then injected into the MoLLE (Molybdenum Liquid-Liquid Extraction) flowsheet, where the Mo is selectively extracted by an organophosphorous acid extractant such as di-(2-ethylhexyl phosphoric acid) (HDEHP). Trace amounts of Nb, Zr, Np, I, and Te are co-extracted. The Mo is stripped using acetohydroxamic acid (AHA) and fed onto an anion exchange column. Following a series of hydroxide, HCl, and oxalic acid wash steps, the Mo can be recovered in NaOH/NaCl. A manuscript describing this process chemistry in more detail was recently published.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Towards the Stable Chelation of Radioantimony(V) for Targeted Auger Theranostics

Abstract Antimony‐119 ( 119 Sb) is one of the most attractive Auger‐electron emitters identified to date, but it remains practically unexplored for targeted radiotherapy because no chelators have been identified to stably bind this metalloid in vivo. In a departure from current studies focused on chelator development for Sb(III), we explore the chelation chemistry of Sb(V) using the tris‐catecholate ligand TREN‐CAM. Through a combination of radiolabeling, spectroscopic, solid‐state, and computational studies, the radiochemistry and structural chemistry of TREN‐CAM with 1XX/nat Sb(V) were established. The resulting [ 1XX Sb]Sb–TREN‐CAM complex remained intact for several days in human serum, signifying high stability under biological conditions. Finally, the first in vivo single photon emission computed tomography and positron emission tomography imaging studies were carried out using 117 Sb, the diagnostic analogue of 119 Sb. These studies revealed marked differences in the uptake and distribution of activity in mice administered unchelated [ 117 Sb]Sb(OH) 6 – versus [ 117 Sb]Sb–TREN‐CAM, suggesting that 117 Sb is largely retained by TREN‐CAM over the time course of the study. Collectively, these findings demonstrate the most physiologically stable complex of no‐carrier‐added 1XX Sb yet reported, offering new promise for the clinical implementation of radioantimony in nuclear medicine. Our results also establish the feasibility of 117 Sb as an elementally matched partner to 119 Sb for theranostic applications.

Olson, Aeli P. [Department of Medical Physics Univ↗
DOE OSTI2020

Report on G4-Med, a Geant4 benchmarking system for medical physics applications developed by the Geant4 Medical Simulation Benchmarking Group

Geant4 is a Monte Carlo code extensively used in medical physics for a wide range of applications, such as dosimetry, micro- and nanodosimetry, imaging, radiation protection, and nuclear medicine. Geant4 is continuously evolving, so it is crucial to have a system that benchmarks this Monte Carlo code for medical physics against reference data and to perform regression testing. In this work, to respond to these needs, we developed G4-Med, a benchmarking and regression testing system of Geant4 for medical physics. G4-Med currently includes 18 tests. They range from the benchmarking of fundamental physics quantities to the testing of Monte Carlo simulation setups typical of medical physics applications. Both electromagnetic and hadronic physics processes and models within the prebuilt Geant4 physics lists are tested. The tests included in G4-Med are executed on the CERN computing infrastructure via the use of the geant-val web application, developed at CERN for Geant4 testing. The physical observables can be compared to reference data for benchmarking and to results of previous Geant4 versions for regression testing purposes. This paper describes the tests included in G4-Med and shows the results derived from the benchmarking of Geant4 10.5 against reference data.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

Dosimetric and biological impact of activity extravasation of radiopharmaceuticals in PET imaging

The increasing use of nuclear medicine and PET imaging has intensified scrutiny of radiotracer extravasation. To our knowledge, this topic is understudied but holds great potential for enhancing our understanding of extravasation in clinical PET imaging. This work aims to (1) quantify the absorbed doses from radiotracer extravasation in PET imaging, both locally at the site of extravasation and with the extravasation location as a source of exposure to bodily organs and (2) assess the biological ramifications within the injection site at the cellular level. A radiation dosimetry simulation was performed using a whole-body 4D Extended Cardiac-Torso (XCAT) phantom embedded in the GATE Monte Carlo platform. A 10-mCi dose of 18 F-FDG was chosen to simulate a typical clinical PET scan scenario, with 10% of the activity extravasated in the antecubital fossa of the right arm of the phantom. The extravasation volume was modeled as a 5.5 mL rectangle in the hypodermal layer of skin. Absorbed dose contributions were calculated for the first two half-lives, assuming biological clearance thereafter. Dose calculations were performed as absorbed doses at the organ and skin levels. Energy deposition was simulated both at the local extravasation site and in multiple organs of interest and converted to absorbed doses based on their respective masses. Each simulation was repeated ten times to estimate Monte Carlo uncertainties. Biological impacts on cells within the extravasated volume were evaluated by randomizing cells and exposing them to a uniform radiation source of 18 F and 68 Ga. Particle types, their energies, and direction cosines were recorded in phase space files using a separate Geant4 simulation to characterize their entry into the nucleus of the cellular volume. Subsequently, the phase space files were imported into the TOPAS-nBio simulation to assess the extent of DNA damage, including double-strand breaks (DSBs) and single-strand breaks (SSBs). Organ-level dosimetric estimations are presented for 18 F and 68 Ga radionuclides in various organs of interest. With 10% extravasation, the hypodermal layer of the skin received the highest absorbed dose of 1.32 ± 0.01 Gy for 18 F and 0.99 ± 0.01 Gy for 68 Ga. The epidermal and dermal layers received absorbed doses of 0.07 ± 0.01 Gy and 0.13 ± 0.01 Gy for 18 F, and 0.14 ± 0.01 Gy and 0.29 ± 0.01 Gy for 68 Ga, respectively. In the extravasated volume, 18 F caused an average absorbed dose per nucleus of 0.17 ± 0.01 Gy, estimated to result in 10.58 ± 0.50 DSBs and 268.11 ± 12.43 SSBs per nucleus. For 68 Ga, the absorbed dose per nucleus was 0.11 ± 0.01 Gy, leading to an estimated 6.49 ± 0.34 DSBs and 161.24 ± 8.12 SSBs per nucleus. Absorbed doses in other organs were on the order of micro-gray (µGy). The likelihood of epidermal erythema resulting from extravasation during PET imaging is low, as the simulated absorbed doses to the epidermis remain below the thresholds that trigger such effects. Moreover, the organ-level absorbed doses were found to be clinically insignificant across various simulated organs. The minimal DNA damage at the extravasation site suggests that long-term harm, such as radiation-induced carcinogenesis, is highly unlikely.

DNA strand breaks↗
DOE OSTI2021

A Comprehensive Review of Selected Major Categories of Lithium Isotope Separation Techniques

The ability to generate materials with an enriched isotopic abundance, that is one that differs from natural abundance, is critical in terms of technology. Isotopes are used in a wide range of applications, including radioactive tracers, nondestructive tests, radiotherapy in human medicine, nuclear fuels, and isotope‐substituted compounds for chemistry and biology research, as well as environmental geochemical signature tracers. A commonly discussed research topic is the isotopic separation of various elements—uranium, hydrogen, lithium, and iodine, among many others. Among these isotopes, lithium isotopes are especially important in various applications including strategic areas. These isotopes are light and hence their separation techniques are not very obvious. Herein, all the main categories of Li isotope separation are reviewed in depth with a focus on electrochemical technologies for stable lithium isotope separation.

Murali, Arun↗
DOE OSTI2025

Global lessons: a comparative analysis of nuclear gamification's impact in American and British students

The global nuclear skills shortage demands urgent educational investment, particularly for students aged 11–18. To help address this, we have introduced RAD Ratings, an interactive card game designed to improve the teaching of nuclear concepts to students, which includes subjects such as half-lives, types of radiation, and radionuclide applications. RAD Ratings has already been successfully piloted in the United Kingdom and in this paper, we present a comparison in selected American schools for the first time. Snapshot survey results show RAD Ratings improve interest in nuclear science and careers, offering insights into gamification’s potential to improve engagement across different geographic locations.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Real-time gross actinide monitoring in extreme (10 6 :1) beta-gamma background fluid streams using mini-ATMFDs

Real-time, trace alpha activity monitoring in extreme mixed radiation fields is a well-known challenge in fission product-rich streams containing mixtures of beta-gamma-alpha radiation-emitting isotopes, arising due to the extreme 10 6 :1 ratio of background beta-gamma vs. alpha radiation activity levels. This paper presents results from a novel approach to addressing this challenge, based on Tensioned Metastable Fluid Detector (TMFD) sensing technology. An acoustically driven technique was developed and tested at Purdue University (using dissolved Rn-Po and U-chain alpha emitters), and validated (with dissolved Pu) at Savannah River National Laboratory for real-time monitoring of alpha activity in the 100–10,000 dpm/mL range.

Environmental Monitoring↗
DOE OSTI2024

Cross-sections for 43 Sc, 44 m Sc, and 44 g Sc from two heavy ion reactions

Two different heavy ion reactions were used to produce 43 Sc (t$_{\frac12}$ = 3.891 h), 44g Sc (t$_{\frac12}$ = 4.042 h), and 44m Sc (t$_{\frac12}$ = 58.61 h) among other stable or long-lived chemically separable products. Production cross sections for 19 F + 27 Al and the reverse kinematic reaction 35 Cl + nat B were measured using an MC-SNICS ion source and the Notre Dame FN Tandem Accelerator. 19 F beams from 35 to 60 MeV were produced with beam currents between 40–80 pnA and 35 Cl beams were produced at six entrance energies with comparable beam currents. This work reports nuclear reaction cross sections 27 Al ( 19 F, x) 43 Sc, 27 Al ( 19 F, pn) 44g Sc, and 27 Al ( 19 F, pn) 44m Sc at six energies between 35 and 60 MeV lab energy. Cross sections within the same energy range were measured for 27 Al ( 19 F, 3pn) 42 K and 27 Al ( 19 F, 3p) 43 K. Comparative measurements were performed for the same compound nucleus produced from nat B( 35 Cl, x) 43 Sc, nat B( 35 Cl, pn) 44g Sc, and nat B( 35 Cl, pn) 44m Sc. The measured thin target cross sections show an overestimation by several statistical models for the scandium radioisotopes. This is corroborated by the measured thick target production rates for both entrance channels. This may be due to angular momentum effects of a heavy ion entrance channel compared to light-ion production, but additional work is required to understand this discrepancy. Finally, these measurements demonstrate that the medically useful 43 Sc, 44g Sc, and 44m Sc radioisotopes can be free of the long-lived contaminant 46 Sc without the use of enriched targets, using heavy ion beams and robust target materials.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

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

The alpha-emitter 225 Ac (t 1/2 = 9.92 d) is currently under development for targeted alpha-particle therapy of cancer, and accelerator production of 225 Ac via proton irradiation of thorium targets requires robust separations of 225 Ac from chemically similar fission product lanthanides. Additionally, the lanthanide elements represent critical components in modern technologies, and radiolanthanides such as 140 Nd (t 1/2 = 3.37 d) also have potential application in the field of nuclear medicine. The ionic liquid, 1-butyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([Bmim][NTf 2 ]), 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 225 Ac and lanthanides. The equilibrium distribution coefficients (Kd) of the rare earth elements (Sc(III), Y(III), Ln(III)), 225 Ac(III), Th(IV), and U(VI) on the prepared DODGAA/[Bmim][NTf 2 ]-SIR were determined from batch adsorption experiments in HCl and HNO 3 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][NTf 2 ]-SIR were effective at separating the lighter lanthanides from each other, and separating 225 Ac from a mixture of lanthanides, 213 Bi, and 225 Ra without the need for additional complexing agents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Development of a SnO 2 -based 44 Ti/ 44 Sc generator for medical applications

Towards application of 44 Sc for diagnostic nuclear medicine, a 44 Ti/ 44 Sc generator based on an inorganic resin has been evaluated. Unlike other radionuclide generators used for medical applications, the long-term retention of the parent 44 Ti is vital due to its long half life. In this work, tin dioxide (SnO 2 ), a robust inorganic-based resin, has been synthesized and used as the stationary phase for a 44 Ti/ 44 Sc generator. The sorption behavior of 44 Ti/ 44 Sc was tested on SnO 2 with varying acids, concentrations, and times. Preliminary batch study results showed >88 % 44 Ti retention to the resin at lower acid concentrations (0.05 M HNO 3 and 0.05 M HCl). A pilot generator was evaluated for a year, demonstrating 85.3 ± 2.8 % 44 Sc elution yields and 0.71 ± 0.14 % 44 Ti breakthrough in 5 M HNO 3 . Based on capacity studies, a 7.4 MBq (200 µCi) upscaled generator system was constructed for further evaluation of the SnO 2 resin stability and the efficacy of the eluted 44 Sc for radiolabeling. 44 Sc could be regularly eluted from this generator in 5 M HNO 3 with an overall average radiochemical yield 84.7 ± 9.5 %. Post-elution processing of the 44 Sc with DGA-normal resin removed all 44 Ti present and allowed for high 44 Sc-DOTA labeling yields of 94.2 ± 0.5 %. Overall, SnO 2 has been shown to be a viable material for a 44 Ti/ 44 Sc generator.

07 ISOTOPE AND RADIATION SOURCES↗