Medical radionuclides such as Ga-68, Cu-64 or Ac-225 are usually produced by irradiation of enriched target materials in cyclotrons or nuclear reactors. After irradiation, the radionuclides need to be separated from their target. While this is mostly done by ion-exchange chromatography, an emerging separation method includes the use of (microfluidic) solvent extraction. However, the extent to which the chelators and organic solvents used during solvent extraction contaminate the final radionuclide-containing solution, including their potential impact on subsequent radiolabeling applications, has not been studied in detail. In this study, the potential contaminants N-benzoyl-N-phenylhydroxilamine (BPHA), dithizone (DIZ) and di(2-ethylhexyl)phosphoric acid (D2EHPA) were investigated, and a microcolumn purification method is proposed. It was found that contaminations with two of these chelators, BPHA and DIZ, significantly interfered with DOTA (1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid) labeling. The applied microcolumn purification method eliminated the BPHA contamination from the Ga-68 solution completely, while simultaneously drastically reducing the total volume and acidity of the solution. It is therefore a promising purification method that can be included in an automated microfluidic solvent extraction procedure.

Background: The radionuclide Ga-68 is commonly used in nuclear medicine, specifically in positron emission tomography (PET). Recently, the interest in producing Ga-68 by cyclotron irradiation of [⁶⁸Zn]Zn nitrate liquid targets is increasing. However, current purification methods of Ga-68 from the target solution consist of multi-step procedures, thus, leading to a significant loss of activity through natural decay. Additionally, several processing steps are needed to recycle the costly, enriched target material. Results: To eventually allow switching from batch to continuous production, conventional batch extraction and membrane-based microfluidic extraction were compared. In both approaches, Ga-68 was extracted using N-benzoyl-N-phenylhydroxylamine in chloroform as the organic extracting phase. Extraction efficiencies of up to 99.5% ± 0.6% were achieved within 10 min, using the batch approach. Back-extraction of Ga-68 into 2 M HCl was accomplished within 1 min with efficiencies of up to 94.5% ± 0.6%. Membrane-based microfluidic extraction achieved 99.2% ± 0.3% extraction efficiency and 95.8% ± 0.8% back-extraction efficiency into 6 M HCl. When executed on a solution irradiated with a 13 MeV cyclotron at TRIUMF, Canada, comparable efficiencies of 97.0% ± 0.4% were achieved. Zn contamination in the back-extracted Ga-68 solution was found to be below 3 ppm. Conclusions: Microfluidic solvent extraction is a promising method in the production of Ga-68 achieving high efficiencies in a short amount of time, potentially allowing for direct target recycling. Graphical Abstract: [Figure not available: see fulltext.].

We present the production of ^58mCo on a small, 13 MeV medical cyclotron utilizing a siphon style liquid target system. Different concentrated iron(III)-nitrate solutions of natural isotopic distribution were irradiated at varying initial pressures and subsequently separated by solid phase extraction chromatography. The radio cobalt (^58m/gCo and ⁵⁶Co) was successfully produced with saturation activities of (0.35 ± 0.03) MBq μA⁻¹ for ^58mCo with a separation recovery of (75 ± 2) % of cobalt after one separation step utilizing LN-resin.

Meitner-Auger-electron emitters have a promising potential for targeted radionuclide therapy of cancer because of their short range and the high linear energy transfer of Meitner-Auger-electrons (MAE). One promising MAE candidate is 197m/gHg with its half-life of 23.8 » h and 64.1 » h, respectively, and high MAE yield. Gold nanoparticles (AuNPs) that are labelled with 197m/gHg could be a helpful tool for radiation treatment of glioblastoma multiforme when infused into the surgical cavity after resection to prevent recurrence. To produce such AuNPs, 197m/gHg was embedded into pristine AuNPs. Two different syntheses were tested starting from irradiated gold containing trace amounts of 197m/gHg. When sodium citrate was used as reducing agent, no 197m/gHg labelled AuNPs were formed, but with tannic acid, 197m/gHg labeled AuNPs were produced. The method was optimized by neutralizing the pH (pH = 7) of the Au/197m/gHg solution, which led to labelled AuNPs with a size of 12.3 ± 2.0 » nm as measured by transmission electron microscopy. The labelled AuNPs had a concentration of 50 » μg (gold)/mL with an activity of 151 ± 93 » kBq/mL (197gHg, time corrected to the end of bombardment).

Medical isotope production of ¹¹C is commonly performed in gaseous targets. The power deposition of the proton beam during the irradiation decreases the target density due to thermodynamic mixing and can cause an increase of penetration depth and divergence of the proton beam. In order to investigate the difference how the target-body length influences the operation conditions and the production yield, a 12 cm and a 22 cm Nb-target body containing N₂/O₂ gas were irradiated using a 13 MeV proton cyclotron. It was found that the density reduction has a large influence on the pressure rise during irradiation and the achievable radioactive yield. The saturation activity of [¹¹C]CO₂ for the long target (0.083 Ci/μA) is about 10% higher than in the short target geometry (0.075 Ci/μA).

In proton therapy, the dose from secondary neutrons to the patient can contribute to side effects and the creation of secondary cancer. A simple and fast detection system to distinguish between dose from protons and neutrons both in pretreatment verification as well as potentially in vivo monitoring is needed to minimize dose from secondary neutrons. Two 3 mm long, 1 mm diameter organic scintillators were tested for candidacy to be used in a proton–neutron discrimination detector. The SCSF-3HF (1500) scintillating fibre (Kuraray Co. Chiyoda-ku, Tokyo, Japan) and EJ-260 plastic scintillator (Eljen Technology, Sweetwater, TX, USA) were irradiated at the TRIUMF Neutron Facility and the Proton Therapy Research Centre. In the proton beam, we compared the raw Bragg peak and spread-out Bragg peak response to the industry standard Markus chamber detector. Both scintillator sensors exhibited quenching at high LET in the Bragg peak, presenting a peak-to-entrance ratio of 2.59 for the EJ-260 and 2.63 for the SCSF-3HF fibre, compared to 3.70 for the Markus chamber. The SCSF-3HF sensor demonstrated 1.3 times the sensitivity to protons and 3 times the sensitivity to neutrons as compared to the EJ-260 sensor. Combined with our equations relating neutron and proton contributions to dose during proton irradiations, and the application of Birks’ quenching correction, these fibres provide valid candidates for inexpensive and replicable proton-neutron discrimination detectors.

In recent years, proton therapy has gained importance as a cancer treatment modality due to its conformality with the tumor and the sparing of healthy tissue. However, in the interaction of the protons with the beam line elements and patient tissues, potentially harmful secondary neutrons are always generated. To ensure that this neutron dose is as low as possible, treatment plans could be created to also account for and minimize the neutron dose. To monitor such a treatment plan, a compact, easy to use, and inexpensive dosimeter must be developed that not only measures the physical dose, but which can also distinguish between proton and neutron contributions. To that end, plastic optical fibers with scintillation materials (Gd₂O₂S:Tb, Gd₂O₂S:Eu, and YVO₄:Eu) were irradiated with protons and neutrons. It was confirmed that sensors with different scintillation materials have different sensitivities to protons and neutrons. A combination of these three scintillators can be used to build a detector array to create a biological dosimeter.

New production routes for ⁹⁹Mo are steadily gaining importance. However, the obtained specific activity is much lower than currently produced by the fission of U-235. To be able to supply hospitals with ⁹⁹Mo/^99mTc generators with the desired activity, the adsorption capacity of the column material should be increased. In this paper we have investigated whether the gas phase coating technique Atomic Layer Deposition (ALD), which can deposit ultra-thin layers on high surface area materials, can be used to attain materials with high adsorption capacity for ⁹⁹Mo. For this purpose, ALD was applied on a silica-core sorbent material to coat it with a thin layer of alumina. This sorbent material shows to have a maximum adsorption capacity of 120 mg/g and has a^99mTc elution efficiency of 55 ± 2% based on 3 executive elutions.