Engineering topics

Brossard, Thomas

Publications and source records attributed to Brossard, Thomas.

6 records

DOE OSTI2023

Accelerator-Based Production of Scandium Radioisotopes for Applications in Prostate Cancer: Toward Building a Pipeline for Rapid Development of Novel Theranostics

In the field of nuclear medicine, the β + -emitting 43 Sc and β - -emitting 47 Sc are promising candidates in cancer diagnosis and targeted radionuclide therapy (TRT) due to their favorable decay schema and shared pharmacokinetics as a true theranostic pair. Additionally, scandium is a group-3 transition metal (like 177 Lu) and exhibits affinity for DOTA-based chelators, which have been studied in depth, making the barrier to implementation lower for 43/47 Sc than for other proposed true theranostics. Before 43/47 Sc can see widespread pre-clinical evaluation, however, an accessible production methodology must be established and each isotope’s radiolabeling and animal imaging capabilities studied with a widely utilized tracer. As such, a simple means of converting an 18 MeV biomedical cyclotron to support solid targets and produce 43 Sc via the 42 Ca(d,n) 43 Sc reaction has been devised, exhibiting reasonable yields. The Nat Ti(γ,p) 47 Sc reaction is also investigated along with the successful implementation of chemical separation and purification methods for 43/47 Sc. The conjugation of 43/47 Sc with PSMA-617 at specific activities of up to 8.94 MBq/nmol and the subsequent imaging of LNCaP-ENZaR tumor xenografts in mouse models with both 43/47 Sc-PSMA-617 are also presented.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2023

Photonuclear production of 47 Ca for 47 Ca/ 47 Sc generator from natural CaCO 3 targets

Here this work investigated the indirect production of 47 Sc from natural Ca targets via 48 Ca(γ,n) 47 Ca → 47 Sc +β - +$\bar{\nu}_e$ with incident electron energies of 30, 35, and 40 MeV. The 47 Ca production yields were simulated using the PHITS Monte Carlo simulation code and compared to experimental data. The simulated production rates for all three irradiations are in good agreement with experimental data within uncertainties. As a demonstration of the 47 Ca/ 47 Sc generator system, one of the irradiated CaCO 3 targets was dissolved in nitric acid, and 47 Sc was isolated from the target material using commercially available Eichrom DGA resin. The 47 Sc was allowed to grow in, and the purification process was repeated with promising 47 Sc and Ca recovery yields.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Fission Induced Radiolysis of Uranyl Sulfate Solutions

At Argonne National Laboratory, an experimental pilot facility is being tested for the production of the medical isotope molybdenum-99 ( 99 Mo). AMORE (Argonne Molybdenum Research Experiment) is an experimental pilot facility in which a 40 MeV electron beam from the LINAC strikes a depleted uranium (DU) target. The target generates high energy neutrons, which are thermalized in an 18-liter uranyl sulfate solution of low enriched uranium (LEU). Thermalized neutrons fission the uranium-235 ( 235 U) to produce 99 Mo as well as many other fission products. The 99 Mo is recovered from the uranium solution and purified. The goal of our work is to reduce the use of highly enriched uranium (HEU) for the production of this isotope.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Half-life of 67 Cu

The half-life of 67 Cu was determined through serial gamma-ray spectrometry measurements of the dominant gamma emission (E γ : 184.6 keV; branching ratio: 48.7%) produced following β- decay. Data were collected consecutively for 1000 s per measurement, with a total of 3063 measurements over the duration of 36 days. The incidence rate for the 184.6 keV gamma-ray was determined from the spectral peak area and duration of each measurement. This rate was then corrected to account for detector dead-time, radioactive decay during each acquisition and drift in the computer clock in comparison to NIST nuclear clock. Least-squares regression analysis was performed to determine the half-life of 67 Cu. The result was 61.761 ± 0.004 h, which is the highest precision measurement to date, and marks a 24-fold precision improvement over the current Nuclear Data Sheets value.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2021

Imaging and dosimetric characteristics of 67 Cu

In recent years the use of beta-emitting radiopharmaceuticals for cancer therapy has expanded rapidly following development of therapeutics for neuroendocrine tumors, prostate cancer, and other oncologic malignancies. One emerging beta-emitting radioisotope of interest for therapy is 67 Cu (t 1/2 : 2.6 d) due to its chemical equivalency with the widely-established positron-emitting isotope 64 Cu (t 1/2 : 12.7 h). Here we evaluate both the imaging and dosimetric characteristics of 67 Cu, as well as producing the first report of SPECT/CT imaging using 67 Cu. To this end, 67 Cu was produced by photon-induced reactions on isotopically-enriched 68 Zn at the Low-Energy Accelerator Facility (LEAF) of Argonne National Laboratory, followed by bulk separation of metallic 68 Zn by sublimation and radiochemical purification by column chromatography. Gamma spectrometry was performed by efficiency-calibrated high-purity germanium (HPGe) analysis to verify absolute activity calibration and establish radionuclidic purity. Absolute activity measurements corroborated manufacturer-recommended dose-calibrator settings and no radionuclidic impurities were observed. Using the Clinical Trials Network anthropomorphic chest phantom, SPECT/CT images were acquired. Medium Energy (ME) SPECT collimation was found to provide the best image quality from the primary 185 keV gamma emission of 67 Cu. Reconstructed images of 67 Cu were similar in quality to images acquired using 177 Lu. Recovery coefficients were calculated and compared against quantitative images of 99m Tc, 177 Lu, and 64 Cu within the same anthropomorphic chest phantom. Production and clinical imaging of 67 Cu appears feasible, and future studies investigating the therapeutic efficacy of 67 Cu-based radiopharmaceuticals are warranted.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2020

Conversion of Uranium Metal to Uranyl Sulfate Solution

Experiments involving conversion of uranium (U) metal to U 3 O 8 , followed by dissolution to form uranyl sulfate solution, were performed with batches of up to 300 g of U metal. The focus was to identify major U species formed during the oxidation process, determine the particle size distribution of the uranium oxide product, and identify the size of the dissolution vessel needed to accommodate the exothermic reaction during the conversion of U 3 O 8 to uranyl sulfate using 30% hydrogen peroxide and H 2 SO 4 . The information obtained in this study will be used for designing full-scale processes to convert kilogram quantities of U metal to uranyl sulfate solution with a concentration of 300 g U/L.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗