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Despite its insufficient availability, Copper-67 is currently attracting much attention for its enormous potential for cancer therapy as theranostic radionuclide. This work aims to accurately measure the unexplored cross section 70Zn(p,x)67Cu in the energy range 45–70 MeV and to evaluate its potential advantages in the case of high-intensity proton beams provided by compact cyclotrons. Thin target foils of enriched 70Zn were manufactured by lamination at the INFN-LNL and irradiated at the ARRONAX facility using the stacked-foils method. A radiochemical procedure for the separation of Cu, Ga and Zn contaminants and the isolation of 67Cu from the irradiated material was developed. The efficiency of the chemical processing was determined for each foil by monitoring the activity of selected tracer radionuclides (61Cu, 66Ga and 69mZn) through γ-spectrometry. Experimental data of the 70Zn(p,x)67Cu, 64Cu, 67Ga, 66Ga, 69mZn, 65Zn cross sections were measured for the first time in the energy range 45–70 MeV and compared with the theoretical results obtained by using the TALYS code. The 67Cu production yield by using enriched 70Zn thick targets was compared with the results obtained by using 68Zn targets in the same irradiation conditions.

Aims: We sought to investigate the thrombogenicity of different DES and BMS in an in vitro system of stent perfusion. Material and methods: The experimental model consisted of a peristaltic pump connected to 4 parallel silicone tubes in which different stents were deployed. Blood was drawn from healthy volunteers and the amount of stent surfaced-induced thrombus deposition was determined using 125I-fibrinogen. Results: Compared to Resolute, Biomatrix and Vision, Xience was associated with the lowest amount of stent surface-induced thrombus formation, with a significant difference compared to Vision (125I-fibrinogen median value deposition [IQ range]: 50 ng [25-98] versus 560 ng [320-1520], respectively, p < 0.05), but not to other DES. In the second set of experiments Fluoropolymer-coated BMS not eluting drug was associated with a significant 3-fold reduction in 125I-fibrinogen deposition (245 ng [80-300]) compared to Vision (625 ng [320-760], p < 0.05), but a 7-fold increase compared to Xience (35 ng [20-60], p < 0.05). Finally Xience was associated with a significantly greater absorption of albumin compared to BMS. Conclusions: In an in vitro system of stent perfusion, Xience was associated with the lowest amount of stent surface-induced thrombus formation compared with Resolute, Biomatrix and Vision, with a noted synergistic effect between the fluoropolymer and the drug.

Background The stability of precursors and reagents is of utmost importance for developing a robust radiolabelling method that provides high and constant radiochemical yield and radiochemical purity. While performing the QC of the [68Ga]Ga-PSMA-11 injectable solutions according to Ph. Eur. Monograph that has recently been published, a trend to the instability of the standard PSMA-11, the same used as a precursor for [68Ga]Ga-PSMA-11 radiosynthesis, has been observed. This instability led to the formation of a side product in a time-dependent manner. The formation of this compound, besides making the implementation of the Ph. Eur. analytical method more difficult, negatively influenced the radiochemical yield and the radiochemical purity by increasing gallium-68 in colloidal and ionic forms. Results The nature of the side product was investigated by adding chelators, such as EDTA, to PSMA-11 solutions and using the combination of UHPLC-HRMS. The results led to the definition of the side product structure, as natFe-PSMA-11, from the combination of the high-affinity chelator HBED-CC, present in the molecule of PSMA-11, and environmental Fe (III). Conclusions Strategies to reduce the risk of low radiolabeling yields and to increase the stability of the PSMA-11 in an aqueous solution were also discussed.

The number of patients with cardiac implantable electronic devices (CIEDs) requiring radiation therapy (RT) for cancer treatment is increasing. The purpose of this study is to estimate the prevalence, possible predictors, and clinical impact of RT-related CIEDs malfunctions. We retrospectively reviewed the medical records of all pacemaker (PM)/implantable cardioverter-defibrillator (ICD) patients who underwent RT in the last 14 years. One hundred and twenty-seven patients who underwent 150 separate RT courses were analysed (99 with a PM and 27 with an ICD). Of note, 21/127 (16.6%) patients were PM-dependent. Neutron-producing RT was used in 37/139 (26.6%) courses, whereas non-neutron-producing RT was used in 102/139 (73.4%) courses. The cumulative dose (Dmax) delivered to the CIED exceeded 5 Gy only in 2/132 (1.5%) cases. Device malfunctions were observed in 3/150 (2%) RT courses, but none was life-threatening or led to a major clinical event and all were resolved by CIED reprogramming. In all cases, the Dmax delivered to the CIED was

During the operational life of a PET cyclotron, the concrete walls of the vault are activated by secondary neutrons. For the dismantling of such accelerator facilities, a considerable amount of low level solid radioactive waste has to be characterized and disposed. To decrease future dismantling costs and complexity, the amount of radioactive waste has to been prospectively evaluated in the design phase, then confirmed at the time of planning decommissioning. In this work, the Monte Carlo code Fluka was used for the assessment of the activation of the bunkers of two different facilities: the 16.5 MeV GE PETtrace at S. Orsola-Malpighi Hospital in Bologna and the 18 MeV IBA Cyclone 18/18 HC at the Bern University Hospital (Inselspital). The simulations were validated by means of experimental measurements performed in our previous works: non-destructive, in field measurements using a portable CZT detector were performed in Bologna; while core drilling samples were extracted from the bunker and measured in laboratory with an HPGe detector in Bern. The activity of the most important radionuclides in the concrete walls of the bunker, namely Eu-152, Mn-54, Co-60, Sc-46, Zn-65 and Cs-134 resulted within the range of 0.01 – 2 Bq/g. The consistency between Monte Carlo results and experimental measurements was within a factor 2 - 3 for most radionuclides, except for Eu-152, Sc-46, Zn-65. The activity concentrations estimated at each position considered exceeds the clearance levels of the new Directive 2013/59/Euratom. The results of this work demonstrate that Monte Carlo simulations based on FLUKA are adequate to assess the residual activation levels, a fundamental information to foresee, plan and optimize the decommissioning of a cyclotron based PET centers.

An assessment study involving the use of the transglutaminase (TGase) conjugation method and the nitride-technetium-99m labelling on a bis(thiosemicarbazone) (BTS) bifunctional chelating agent is presented. The previously described chelator diacetyl-2-(N-4-methyl-3-thiosemicarbazone)-3-(N-4-amino-3-thiosemicarbazone), H(2)ATSM/A, has been functionalized with 6-aminohexanoic acid (epsilon-Ahx) to generate the bifunctional chelating agent diacetyl-2-(N-4-methyl-3-thiosemicarbazone)-3-[N-4-(amino)-(6-aminohexanoic acid)-3-thiosemicarbazone], H(2)ATSM/A-epsilon-Ahx (1), suitable for conjugation to glutamine (Gln) residues of bioactive molecules via TGase. The feasibility of the TGase reaction in the synthesis of a bioconjugate derivative was investigated using Substance P (SP) as model peptide. Compounds 1 and H(2)ATSM/A-epsilon-Ahx-SP (2) were labelled with nitride-technetium-99m, obtaining the complexes [Tc-99m] [Tc(N)(ATSM/A-epsilon-Ahx)] ((99m)Tc1)and [Tc-99m][Tc(N)(ATSM/ A-epsilon-Ahx-SP)] ((99m)Tc2). The chemical identity of (99m)Tc1 and (99m)Tc2 was confirmed by radio/UV-RP-HPLC combined with ESI-MS analysis on the respective carrier-added products (99g/99m)Tc1 and (99g/99m)Tc2. The stability of the radiolabelled complexes after incubation in various environments was investigated. All the results were compared with those obtained for the corresponding Cu-64-analogues, (64)Cu1 and (64)Cu2. The TGase reaction allows the conjugation of 1 with the peptide, but it is not highly efficient due to instability of the chelator in the required conditions. The SP-conjugated complexes are unstable in mouse and human sera. However, indeed the BTS system can be exploited as nitride-technetium-99m chelator for highly efficient technetium labelling, thus making compound 1 worthy of further investigations for new targeted technetium and copper radio-pharmaceuticals encompassing Single Photon Emission Computed Tomography and Positron Emission Tomography imaging.

OBJECTIVE: The aim of this study was to analyse early and delayed acquisition on copper-64 diacetyl-bisN4-methylthiosemicarbazone (Cu-ATSM) PET/CT in a small cohort of patients by comparing semiquantitative and computer-aided fractal geometry analyses. PATIENTS AND METHODS: Five cancer patients, including non-small-cell lung cancer and head and neck cancer, were investigated with Cu-ATSM PET/CT. Participants received an intravenous injection of Cu-ATSM according to body size and were imaged 60 min (early) and 16 h (delayed) later on hybrid PET/CT. Reconstructed images were visualized on advanced workstations for the definition of semiquantitative parameters: standardized uptake value (SUV)max, SUVratio-to-muscle, SUVmean, hypoxic volume (HV) and hypoxic burden (HB=HV×SUVmean). DICOM data retrieved from both scans were analysed using an ad-hoc computer program to determine the mean intensity value, SD, relative dispersion, three-dimensional histogram fractal dimension and three-dimensional fractal dimension. RESULTS: All tumour lesions showed increased uptake of Cu-ATSM at early evaluation, with a median SUVratio-to-muscle of 4.42 (range: 1.58-5.62), a median SUVmax of 5.3 (range: 1.9-7.3), a median SUVmean of 2.8 (range: 1.5-3.9), a median HV of 41.6 cm (range: 2.8-453.7) and a median HB of 161.5 cm (range: 4.4-1112.5). All semiquantitative data obtained at 1 h were consistent with the parameters obtained on delayed imaging (P>0.05). A borderline statistically significant difference was found only for SUVmax of the muscle (P=0.045). Fractal geometry analysis on DICOM images showed that all parameters at early imaging showed no statistically significant difference with late acquisition (P>0.05). CONCLUSION: Our findings support the consistency of Cu-ATSM PET/CT images obtained at early and delayed acquisition for the assessment of tumour lesions.

During the operational life of a PET Cyclotron, the concrete walls of its vault are activated by secondary neutrons. For this reason when considering dismantling of such accelerators, a considerable amount of low level solid radioactive waste has to be characterized and disposed. To decrease future dismantling costs, the amount of radioactive waste has to be evaluated in advance to define an optimum decommissioning strategy. Aim of this work is to develop methodologies for the preliminary assessment of activation induced in a bunker via experimental measurements. Two different facilities were studied and different approaches were adopted: non-destructive, in field measurements using a portable CZT detector were performed in the bunker of a 16.5 MeV GE PETtrace (S. Orsola-Malpighi Hospital of Bologna); while core drilling samples were extracted from the bunker of an 18 MeV IBA Cyclone 18 H C (Inselspital in Bern) and measured in laboratory with an HPGe detector. The principal long-lived radionuclides found in concrete were Eu-152, Mn-54, Co-60, Sc-46, Cs-134 with activity concentrations comprised between 0.01 and 0.6 Bq/g. In all performed measurements the total activity concentration exceed the clearance level of Directive (2013)/59/Euratom. Considering a hypothetical decommissiong of the two facilities, the amount of low lever radioactive waste estimated was about 93 m3 in the case of S. Orsola-Malpighi, while about 58 m3 for Insespital facility. Both the methodologies have proved to be adequate for the purposes.

MiRNAs are single stranded RNAs of 18–22 nucleotides. They are promising diagnostic and prognostic markers for several pathologies including tumors, neurodegenerative, cardiovascular and autoimmune diseases. In the present work the development and characterization of anti-miRNA radiolabeled probes based on peptide nucleic acids (PNAs) for potential non-invasive molecular imaging in vivo of giant cell arteritis are described. MiR-146a and miR-146b-5p were selected as targets because they have been found up-regulated in this disease. Anti-miR and scramble PNAs were synthesized and linked to carboxyfluorescein or DOTA. DOTA-anti-miR PNAs were then labelled with copper-64 (64Cu) to function as non-invasive molecular imaging tools. The affinity of the probes for the targets was assessed in vitro by circular dichroism and melting temperature. Differential uptake of fluorescein and 64Cu labeled anti-miRNA probes was tested on BCPAP and A549 cell lines, expressing different levels of miR-146a and -146b-5p. The experiments showed that the anti-miR-146a PNAs were more effective than the anti-miR-146b-5p PNAs. Anti-miR-146a PNAs could bind both miR-146a and miR-146b-5p. The uptake of fluorescein and 64Cu labeled anti-miR-146a PNAs was higher than that of the negative control scramble PNAs in miRNA expressing cells in vitro. 64Cu-anti-miR-146a PNAs might be further investigated for non-invasive PET imaging of miR-146 overexpressing diseases.

Technetium-99m (99mTc) is the most used radionuclide worldwide in nuclear medicine for diagnostic imaging procedures. 99mTc is typically extracted from portable generators containing 99Mo, which is produced normally in nuclear reactors as a fission product of highly enriched Uranium material. Due to unexpected outages or planned and unplanned reactor shutdown, significant 99mTc shortages appeared as a problem since 2008 The alternative cyclotron-based approach through the 100Mo(p,2n)99mTc reaction is considered one of the most promising routes for direct 99mTc production in order to mitigate potential 99Mo shortages. The design and manufacturing of appropriate cyclotron targets for the production of significant amounts of a radiopharmaceutical for medical use is a technological challenge. In this work, a novel solid target preparation method was developed, including sputter deposition of a dense, adherent, and non-oxidized Mo target material onto a complex backing plate. The latter included either chemically resistant sapphire or synthetic diamond brazed in vacuum conditions to copper. The target thermo-mechanical stability tests were performed under 15.6 MeV proton energy and different beam intensities, up to the maximum provided by the available GE Healthcare (Chicago, IL, USA) PET trace medical cyclotron. The targets resisted proton beam currents up to 60 µA (corresponding to a heat power density of about 1 kW/cm2) without damage or Mo deposited layer delamination. The chemical stability of the proposed backing materials was proven by gamma-spectroscopy analysis of the solution obtained after the standard dissolution procedure of irradiated targets in H2O2.

Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of 41Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility.

This work was aimed at characterizing and validating a very compact, USB-powered, CdZnTe detector for gamma-ray spectrometry applications in radiopharmacy. A GR1 model (by KromekTM, Sedgefield, UK) was utilized. The detector was calibrated in energy and efficiency. Samples of [18F]FDG and [68Ga]-DOTANOC were measured to assess the detector’s suitability in radionuclidic purity measurements. The energy resolution and the efficiency obtained are fully adequate for quality control of positron emission tomography radiopharmaceuticals. Results of [18F]FDG and [68Ga]-DOTANOC samples proved to be comparable to those obtained using high-purity Germanium detectors. The performance of the detector and its very compact size make this type of device an extremely attractive tool for QC applications in radiopharmacy.

During the operational life of a PET Cyclotron, the concrete walls of the cyclotron vault are activated by the secondary neutron flux interacting with rare earths and metals present in the concrete or in reinforcement bars. For this reason when considering dismantling of such accelerators, the amount of radioactive waste has to be evaluated in advance to identify any critical issues or possible countermeasures to be taken to define an optimum decommissioning strategy. The aim of this work is to define a non-destructive in situ measurement methodology for a preliminary activation assessment of a cyclotron bunker with no need for core drilling. A very compact, USB-powered, CdZnTe (CZT) detector for gamma-ray spectrometry was used for the activation assessment of the site of installation of a GE PETtrace (16.5 MeV) cyclotron, routinely used in the production of positron-emitting radionuclides. Because of the complexity of measurement geometry, the efficiency calibration of the detector was performed via Monte Carlo (MC) simulations. The detector was accurately modelled using FLUKA, including a 5 cm lead shielding set-up. The MC model of the detector was validated for a wide range of energies and different source geometries, showing discrepancies below 5% for all tested sources. The efficiency curve for wall activation measurements was calculated, allowing a quantitative evaluation of activity concentration.

LARAMED project aims to set up an advanced science and technology facility to develop new and efficient methods to produce medical radioisotopes at Legnaro National laboratories of National Institute of Nuclear Physics. Waiting for the facility full operation, LARAMED group has already started working on the cyclotron production of several conventional and emerging radionuclides. Suitable target preparation is one of the most critical aspects in cyclotron production of radioisotopes. LARAMED group has investigated a set of non-classical techniques for metallic target preparation. Magnetron sputtering technique developed for ultra-thick film deposition was applied for the preparation of the Mo solid targets for 99mTc production, and Y ones for 89Zr production. Spark plasma sintering method provides efficient sintering of powders and good bonding of metallic pellet to a backing. Like magnetron sputtering, it was tested for natl00Mo and natY targets, as well as for natCr targets preparation aimed at 52Mn production. The High energy Vibration Powders Plating technique was instead applied for natural (Mo, Ti) and enriched 48Ti metallic powders with >95% deposition efficiency. All three techniques tested provided the cyclotron solid targets with high thermomechanical performance under the beam (IkW/cm2).

In the last years, the technology for producing the important medical radionuclide technetium-99m by cyclotrons has become sufficiently mature to justify its introduction as an alternative source of the starting precursor [99mTc][TcO4]- ubiquitously employed for the production of 99mTc-radiopharmaceuticals in hospitals. These technologies make use almost exclusively of the nuclear reaction 100Mo(p,2n)99mTc that allows direct production of Tc-99m. In this study, it is conjectured that this alternative production route will not replace the current supply chain based on the distribution of 99Mo/99mTc generators, but could become a convenient emergency source of Tc-99m only for in-house hospitals equipped with a conventional, low-energy, medical cyclotron. On this ground, an outline of the essential steps that should be implemented for setting up a hospital radiopharmacy aimed at the occasional production of Tc-99m by a small cyclotron is discussed. These include (1) target production, (2) irradiation conditions, (3) separation/purification procedures, (4) terminal sterilization, (5) quality control, and (6) Mo-100 recovery. To address these issues, a comprehensive technology for cyclotron-production of Tc-99m, developed at the Legnaro National Laboratories of the Italian National Institute of Nuclear Physics (LNL-INFN), will be used as a reference example.

Purpose In recent years the use of 68Ga (t1/2 = 67.84 min, β+: 88.88%) for the labelling of different PET radiopharmaceuticals has significantly increased. This work aims to evaluate the feasibility of the production of 68Ga via the 68Zn(p,n)68Ga reaction by proton irradiation of an enriched zinc solution, using a biomedical cyclotron, in order to satisfy its increasing demand. Methods Irradiations of 1.7 M solution of 68Zn(NO3)2 in 0.2 N HNO3 were conducted with a GE PETtrace cyclotron using a slightly modified version of the liquid target used for the production of fluorine-18. The proton beam energy was degraded to 12 MeV, in order to minimize the production of 67Ga through the 68Zn(p,2n)67Ga reaction. The product’s activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector. Results The saturation yield of 68Ga amounts to (330 ± 20) MBq/µA, corresponding to a produced activity of 68Ga at the EOB of (4.3 ± 0.3) GBq in a typical production run at 46 µA for 32 min. The radionuclidic purity of the 68Ga in the final product, after the separation, is within the limits of the European Pharmacopoeia (>99.9%) up to 3 h after the EOB. Radiochemical separation up to a yield not lower than 75% was obtained using an automated purification module. The enriched material recovery efficiency resulted higher than 80–90%. Conclusions In summary, this approach provides clinically relevant amounts of 68Ga by cyclotron irradiation of a liquid target, as a competitive alternative to the current production through the 68Ge/68Ga generators.

Background In medical cyclotron facilities, 11C is produced according to the 14N(p,α)11C reaction and widely employed in studies of prostate and brain cancers by Positron Emission Tomography. It is known from literature that the 11C-target assembly shows a reduction in efficiency during time, meaning a decrease of activity produced at the end of bombardment. This effect might depend on aspects which are still not completely known. Objective Possible causes of the loss of performance of the 11C-target assembly were addressed by Monte Carlo simulations. Methods Geant4 was used to model the 11C-target assembly of a GE PETtrace cyclotron. The physical and transport parameters to be used in the energy range of medical applications were extracted from literature data and 11C routine productions. The Monte Carlo assessment of 11C saturation yield was performed varying several parameters such as the proton energy and the angle of the target assembly with respect to the proton beam. Results The estimated 11C saturation yield is in agreement with IAEA data at the energy of interest, while it is about 35% greater than the experimental value. A more comprehensive modeling of the target system, including thermodynamic effect, is required. The energy absorbed in the inner layer of the target chamber was up to 46.5 J/mm2 under typical irradiation conditions. Conclusion This study shows that Geant4 is potentially a useful tool to design and optimize targetry for PET radionuclide productions. Tests to choose the Geant4 physics libraries should be performed before using this tool with different energies and materials.

Accurate determination of calibration factors for radionuclide activity meters is crucial for quantitative studies and in the optimization step of radiation protection, as these detectors are widespread in radiopharmacy and nuclear medicine facilities. In this work we developed the Monte Carlo model of a widely used activity meter, using the Geant4 simulation toolkit. More precisely the "PENELOPE" EM physics models were employed. The model was validated by means of several certified sources, traceable to primary activity standards, and other sources locally standardized with spectrometry measurements, plus other experimental tests. Great care was taken in order to accurately reproduce the geometrical details of the gas chamber and the activity sources, each of which is different in shape and enclosed in a unique container. Both relative calibration factors and ionization current obtained with simulations were compared against experimental measurements; further tests were carried out, such as the comparison of the relative response of the chamber for a source placed at different positions. The results showed a satisfactory level of accuracy in the energy range of interest, with the discrepancies lower than 4% for all the tested parameters. This shows that an accurate Monte Carlo modeling of this type of detector is feasible using the low-energy physics models embedded in Geant4. The obtained Monte Carlo model establishes a powerful tool for first instance determination of new calibration factors for non-standard radionuclides, for custom containers, when a reference source is not available. Moreover, the model provides an experimental setup for further research and optimization with regards to materials and geometrical details of the measuring setup, such as the ionization chamber itself or the containers configuration.

Background Unilateral condylar hyperplasia (UCH) of the mandible, or Hypercondylia, is a pathological condition that determines an abnormal growth of the affected condyle. Bone SPECT with Tc99m-diphosphonates is a successful tool in the diagnosis of UCH. EANM guidelines also suggest the use of 18F–NaF PET/CT, though it leads to a higher radiation exposure. Aim As UCH patients are young, we aimed to develop a low dose 18F–Fluoride PET/CT protocol and compare it to a standard injected activity scan, to assess if the image quality remains unchanged. Materials and methods We prospectively enrolled 20 patients (7 males, 13 females, mean age 23.2) with UCH, who underwent 18F–NaF PET/CT to assess the hypercondylia. We administered a low activity of 18F–NaF (2.9 MBq/kg) in 15 patients and a standard activity (5.3 MBq/kg) in 5 patients. Activity range was chosen according to 2015 EANM guidelines. To determine if the scans with low radiotracer activity were “diagnostic” such as those with standard activity, two expert nuclear medicine physicians, unaware of the administered activity, independently reviewed the scans and expressed a final qualitative judgment in terms of “diagnostic”/“non-diagnostic” scan. Furthermore, we compared the effective dose of a low injected activity PET/CT to the standard one and to a Bone SPECT performed with standard injected activity of Tc99m-diphosphonates. Results Reviewers classified 19 of 20 scans as “diagnostic”. Only one of them was classified as “non diagnostic” due to condylar arthrosis that disturbed the correct evaluation of condylar radiotracer uptake. The effective dose of a 18F–Fluoride PET/CT, in patient of 70 kg, is about 3.5 mSv in scans performed with 2.9 MBq/kg [0.017 mSv/MBq × 2.9 MBq/kg × 70 kg] and about 6.3 mSv in ones performed with 5.3 MBq/kg [0.017 mSv/MBq × 5.3 MBq/kg × 70 kg]. The effective dose of 99mTc-MDP bone SPECT is about 3.2 mSv [0.0043 mSv/MBq × 740 MBq of 99mTc-MDP]. Discussion 18F–NaF PET/CT performed with a low radiotracer activity allows a good assessment of UCH similar to that performed with an ordinary activity. The effective radiation dose of a low-injected activity PET/CT is significantly lower than an ordinary-injected activity and is not significantly higher than the most used Bone SPECT. Moreover PET/CT is performed in 1.5 h while Bone SPECT requires at least 3.5 h. Conclusions The 18F–Fluoride PET/CT procedure could be performed with 2.9 MBq/Kg (minimum 185 MBq, recommended at least 200 MBq) of 18F–NaF to minimize the effective radiation dose received, maintaining the quality of the scan. Further studies including a larger number of patients and clinical follow-up are needed to confirm our preliminary findings.

Purpose The aim of this work was to evaluate the feasibility of production of 68Ga via the 68Zn(p,n)68Ga reaction by proton irradiation of enriched zinc nitrate solution using a biomedical cyclotron, an alternative method to the current production through the 68Ge/68Ga generators. Methods The target material was prepared by dissolving metallic 68Zn (enrichment of 99.26%) in HNO3 6 N and then diluting it with HNO3 0.2 N to obtain a 1.7 M solution of 68Zn(NO3)2 [1] . Using the cross-section data (EXFOR database) and the proton stopping power (SRIM software), the 68Ga saturation yield and its impurities were estimated. Irradiation tests were performed using the GE PETtrace cyclotron, located at our Institution. A liquid target was used, with a 25 μm foil of Havar® and a 250 μm foil of Niobium, employed to obtain proper selection of entrance beam energy, to degrade the energy of the beam to 12 MeV in order to minimize the production of 67Ga through the 68Zn(p,2n)67Ga reaction. The modification of the target also allowed us to improve resistance of the foil to acidity of the target material. Irradiations were carried out for 32 ± 1 min at 47 ± 2 μA. The product’s activity was measured using an activity meter calibrated with a standard of 68Ge traceable to NIST, while radionuclidic purity was evaluated by using gamma-ray spectrometry with an HPGe detector. Results The gamma-spectra of the irradiated samples showed the presence of different isotopes of Ga, Cu and Co, as predicted by the theoretical estimates. Radionuclidic purity is within the limits of EU Pharmacopeia up to 3 h after end of bombardment. Separation and recovery up to a yield of 90 ± 9% was obtained. Conclusions In summary, although optimizations and further advances are necessary, the results obtained are very promising and demonstrate the feasibility to produce sufficient activity of 68Ga in cyclotron as a competitive alternative to the 68Ge/68Ga generators.

Purpose 41Ar (t1/2 = 109.34 m) is produced by the activation of air due to the neutron flux during production of PET nuclides in a medical cyclotron facility, according to the 40Ar(n, γ) 41Ar reaction. We describe a relatively inexpensive and readily reproducible methodology of air sampling used for quantification of 41Ar during routine production of PET nuclides. Methods The activity concentration of 41Ar was evaluated during the routine production of 18 F− performed using a GE PETtrace Cyclotron. Irradiations were performed at (53–60) μA of current on a Niobium body, high pressure target, filled with 2.5 mL of 99% enriched 18O water. We realized an air sampling system that allows collecting representative air samples during an irradiation within the cyclotron bunker, in the ducts of the ventilation system, just before final filtering and before emission to external atmosphere. The air sampling system is composed by polyethylene tube, fitted to a Marinelli beaker (total filling volume 1.41 dm3) and to an air pump; the flow rate was set to 2 dm3/min. For all samples, at the end of the irradiation, each Marinelli beakers was brought to the gamma ray spectrometry laboratory; the beaker was measured using an HPGe detector. In order to assess the Total Effective Dose (TED) to the population living in the area surrounding the facility, the data on the total amount of 41Ar released were input in the code HotSpot. Results A total of 39 samples collected in 4 different conditions were used and analyzed; in particular: inside the cyclotron bunker with ventilation off, inside the cyclotron bunker with ventilation on, in the extraction air duct before and after the final filter. For all samples, the only radionuclide identified was 41Ar. The average 41Ar saturation yield per one litre of air emitted in the environment, after the final filter, resulted to be (0.044 ± 0.007) Bq/(μA·dm3). The maximum value of TED for the critical group of the population resulted less than 0.19 μSv/year. Conclusions Based on the experimental evidence, the emission of 41Ar in a busy PET cyclotron Centre is not radiologically relevant (typically assumed as 10 μSv/year).

Biomedical cyclotrons for production of Positron Emission Tomography (PET) radionuclides and radiotherapy with hadrons or ions are widely diffused and established in hospitals as well as in industrial facilities and research sites. Guidelines for site planning and installation, as well as for radiation protection assessment, are given in a number of international documents; however, these well-established guides typically offer analytic methods of calculation of both shielding and materials activation, in approximate or idealized geometry set up. The availability of Monte Carlo codes with accurate and up-to-date libraries for transport and interactions of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of nowadays computers, makes systematic use of simulations with realistic geometries possible, yielding equipment and site specific evaluation of the source terms, shielding requirements and all quantities relevant to radiation protection. In this work, the well-known Monte Carlo code FLUKA was used to simulate two representative models of cyclotron for PET radionuclides production, including their targetry; and one type of proton therapy cyclotron including the energy selection system. Simulations yield estimates of various quantities of radiological interest, including the effective dose distribution around the equipment, the effective number of neutron produced per incident proton and the activation of target materials, the structure of the cyclotron, the energy degrader, the vault walls and the soil. The model was validated against experimental measurements and comparison with well-established reference data. Neutron ambient dose equivalent H ⁎ (10) was measured around a GE PETtrace cyclotron: an average ratio between experimental measurement and simulations of 0.99±0.07 was found. Saturation yield of 18 F, produced by the well-known 18 O(p,n) 18 F reaction, was calculated and compared with the IAEA recommended value: a ratio simulation to IAEA of 1.01±0.10 was found.

The introduction of 68 Ga labelled tracers has changed the diagnostic approach to neuroendocrine tumours and the availability of a reliable, long-lived 68 Ge/ 68 Ga generator has been at the bases of the development of 68 Ga radiopharmacy. The huge increase in clinical demand, the impact of regulatory issues and a careful radioprotection of the operators have boosted for extensive automation of the production process. The development of automated systems for 68 Ga radiochemistry, different engineering and software strategies and post-processing of the eluate were discussed along with impact of automation with regulations.

The aim of this study was to synthesize and preclinically evaluate anSeveral conditions for the labeling ofQuantitative labeling yields could be achieved with97 % radiochemical purity. The

In the planning of a new cyclotron facility, an accurate knowledge of the radiation field around the accelerator is fundamental for the design of shielding, the protection of workers, the general public and the environment. Monte Carlo simulations can be very useful in this process, and their use is constantly increasing. However, few data have been published so far as regards the proper validation of Monte Carlo simulation against experimental measurements, particularly in the energy range of biomedical cyclotrons. In this work a detailed model of an existing installation of a GE PETtrace 16.5 MeV cyclotron was developed using FLUKA. An extensive measurement campaign of the neutron ambient dose equivalent H ∗ (10) in marked positions around the cyclotron was conducted using a neutron rem-counter probe and CR39 neutron detectors. Data from a previous measurement campaign performed by our group using TLDs were also re-evaluated. The FLUKA model was then validated by comparing the results of high-statistics simulations with experimental data. In 10 out of 12 measurement locations, FLUKA simulations were in agreement within uncertainties with all the three different sets of experimental data; in the remaining 2 positions, the agreement was with 2/3 of the measurements. Our work allows to quantitatively validate our FLUKA simulation setup and confirms that Monte Carlo technique can produce accurate results in the energy range of biomedical cyclotrons.

PURPOSE: Cu-ATSM is a very promising PET radiopharmaceutical for tumor imaging of hypoxia. One of the advantages of this compound compared with other hypoxia-avid tracers is the high tumor-to-background signal offered, which guaranties facilitated tumor delineation. This study analyzes optimal semiquantitative and quantitative parameters obtained by Cu-ATSM PET/CT in the same cohort of patients with special focus on their correlation to disease outcome. PATIENTS AND METHODS: A prospective recruitment of 18 consecutive patients (M:F, 13:5; mean age, 60.7 years) with locally advanced non-small cell lung cancer (n = 7) or head and neck cancer (HNC) was performed. Each participant received 105 to 500 MBq of tracer according to body size and was scanned in a 3-dimensional mode PET/CT 60 minutes after tracer injection. PET images were reconstructed and visualized on a GE Advanced 4.6 workstation for the definition of semiquantitative and quantitative parameters: SUVmax, SUVratio-to-muscle, hypoxic tumor volume (HTV), and hypoxic burden (HB = HTV × SUVmean). These data were subsequently correlated to disease outcome, expressed in terms of progression-free survival calculated on a follow-up period with a median of 14.6 months. RESULTS: All patients showed a moderately to highly increased uptake of Cu-ATSM in tumor lesions, with a mean SUVmax of 5.2 (range, 1.9-8.3) and mean SUVratio of 4.4 (range, 1.6-6.8). In addition, a broad range of HTV and HB was defined as mean values of 99.3 cm (range, 2.5-453.7 cm) and 301 (4.2-1134), respectively. Receiver operating characteristic analysis identified as reference cutoffs with respect to disease outcome with the following values: SUVmax >2.5 (AUC, 0.57; sensitivity, 88.9%; specificity, 50%), SUVratio ≤4.4 (AUC, 0.60; sensitivity, 50; specificity, 83.3%), HTV >160.7 cm (AUC, 0.61; sensitivity, 55.6%; specificity, 75%), and HB >160.7 (AUC, 0.67; sensitivity, 58.3%; specificity, 83.3%). In our cohort, HB showed a statistically significant difference in terms of mean values on the analysis of variance test with respect to disease progression (P = 0.04). On univariate analysis, Cox regression confirmed these findings and showed a significant correlation to progression-free survival for HB (P = 0.05) and HTV (P = 0.02). CONCLUSIONS: In our cohort, the definition of optimal semiquantitative and quantitative parameters on Cu-ATSM PET/CT seems feasible and in line with previously published data. However, when considering the prognostic role with respect to disease outcome, the more robust parameters are represented by HTV and HB.

The design and fabrication of a fully-automated, remotely controlled module for the extraction and purification of technetium-99m (Tc-99m), produced by proton bombardment of enriched Mo-100 molybdenum metallic targets in a low-energy medical cyclotron, is here described. After dissolution of the irradiated solid target in hydrogen peroxide, Tc-99m was obtained under the chemical form of 99mTcO4-, in high radionuclidic and radiochemical purity, by solvent extraction with methyl ethyl ketone (MEK). The extraction process was accomplished inside a glass column-shaped vial especially designed to allow for an easy automation of the whole procedure. Recovery yields were always >90% of the loaded activity. The final pertechnetate saline solution Na99mTcO4, purified using the automated module here described, is within the Pharmacopoeia quality control parameters and is therefore a valid alternative to generator-produced 99mTc. The resulting automated module is cost-effective and easily replicable for in-house production of high-purity Tc-99m by cyclotrons.

Cyclotrons dedicated to the production of PET radiopharmaceuticals for clinical use can be a resource also for research in other fields of science. In the present study we assessed the feasibility of (65)Zn production via the (65)Cu(p,n)(65)Zn reaction using a 16.5 MeV GE-PETtrace biomedical cyclotron. The radioisotope (65)Zn (t(1/2)=244.26 zd; decay mode: EC 98.3%, β(+) 1.7%) has a wide range of applications, especially in botany and agriculture; its long half life allows storage and delivery to research centers without a cyclotron on site. Natural copper foils (thickness: 100 μm; purity: >99.9%) were irradiated in a solid target station developed in our institution. The EXFOR and IAEA cross sections databases were carefully studied to optimize the irradiation setup in order to maximize the yield of the radionuclide of interest and minimize the production of isotopic impurities. We performed production runs at 20 μA for 30 min. Activity was measured using a calibrated HPGe gamma ray detector and an ionization chamber radionuclide activity meter. These measures showed the presence of two main contaminants: (63)Zn (t(1/2)=38.47 min) and (64)Cu (t(1/2)=12.7h); we produced (65)Zn with a saturation yield of (1.06 ± 0.07) GBq/μA (E(p)=12.65→10.48 MeV). The irradiated target was dissolved in 1 ml of 6N HNO(3) at the temperature of 90°C and then purified through anion exchange chromatography with BIO RAD AG 1×8 resin. The radionuclidic purity of the final solution of (65)Zn in HCl 0.005 N, expressed in activity, resulted superior to (99.921 ± 0.003) %.

The widely used Monte Carlo simulation code FLUKA has been utilized to prototype a solid target for the production of 89Zr by irradiation of a metallic 89Y target foil in a 16.5 MeV proton biomedical cyclotron, through the reaction 89Y(p, n)89Zr. Simulations were performed with and without an Al energy degrader. In the setup of the geometry of the target, state of the art support tools, like SimpleGeo, were used for accurate, detailed modeling. The results permitted a quick assessment of all possible radionuclidic contaminants and confirmed that the use of an energy degrader avoids production of the most important impurity, 88Zr. The estimated value for the activity produced in one hour of irradiation at 20 μA is 384±42 MBq; this is encouraging, indicating possible production of clinically significant amounts of activity with the relatively simple target setup adopted. Initial experimental tests gave results in excellent agreement with simulations, confirming the usefulness and accuracy of FLUKA as a tool for the design and optimization of targets for the production of PET radionuclides.

The radionuclide89Zr (T1/2=78.4ߙh) is particularly attractive for in vivo assessment of biochemical proesses characterized by biological half times in the order of several hours. In the present study we assessed the feasibility of89Zr productionviathe89Y(p,n)89Zr reaction. Irradiation tests were performed using a 16.5ߙMeV GE-PETtrace cyclotron. High purity (>99.9%) yttrium metallic foils (0.15ߙmm) were irradiated in a solid target station developed in our Institution. The literature cross section data of all the nuclear reactions possible in the energy range of interest were carefully studied to optimize the irradiation parameters. The irradiated target was dissolved in 1ߙN HCl and the activity of the sample was measured in a CRC-15 PET dose calibrator, setting different values for the calibration factor. The sample activity was then measured usin g a calibrated HPGe gamma ray detector. A comparison of the activity measurements allowed the evaluation of the calibration factor for the Capintec CRC-15 PET; for a 3ߙml syringe geometry, it was found 739±60. Saturation yields of89Zr were found to be 1150±110ߙMBq/μA for a single 0.15ߙmm thick disc irradiation (Ep=12.6→11.2ߙMeV) and 2400±220ߙMBq/μA for irradiation of two discs of total thickness 0.3ߙmm (Ep=12.6→9.5MeV). Total produced activities in the range of 200–400ߙMBq (according to target thickness) were repeatedly and safely obtained by 60ߙmin bombardments at 20ߙμA. Analysis of gamma ray spectra of all the samples showed a very high radionuclidic purity (>99.9995%).

With the ever-increasing number of cyclotron installations, and therefore of the maintenance personnel involved, the possibility of swift, 'yes or no' screening for internal contamination becomes a prized asset. The present work presents one such procedure, evolved from an approximate whole body counting technique in widespread use in emergency situations. A detailed analysis of possible pathways for contamination leads to pinpointing the nuclides of interest. Different calibration methods are applied, showing moderate variation among them. The minimum detectable activity of order 1000 Bq is determined. The method proves sensitive enough to exclude significant contamination, or to identify its presence instantly 'on site' to prompt further in-depth investigation.

Ozone produced from radiolytic processes was investigated as a possible health hazard in the working environment at the University Hospital “S.Orsola—Malpighi” PET facility. Intense radiation fields can generate ozone, known to be the most toxic gas produced by ionizing radiation around a particle accelerator. To evaluate ozone concentration in air, two different measurement campaigns were conducted with passive diffusion detectors. Comparison of the results with the concentration limits recommended by American Conference of Governmental Industrial Hygienists (ACGIH) demonstrated that ozone poses no health hazard to workers around a biomedical cyclotron.

Evaluation of the radioisotopic purity of technetium-99m (99mTc) produced in GBq amounts by proton bombardment of enriched molibdenum-100 (100Mo) metallic targets at low proton energies (i.e. within 15-20 MeV) is conducted. This energy range was chosen since it is easily achievable by many conventional medical cyclotrons already available in the nuclear medicine departments of hospitals. The main motivation for such a study is in the framework of the research activities at the international level that have been conducted over the last few years to develop alternative production routes for the most widespread radioisotope used in medical imaging. The analysis of technetium isotopes and isomeric states (9xTc) present in the pertechnetate saline Na99mTcO4 solutions, obtained after the extraction/purification procedure, reveals radionuclidic purity levels basically in compliance with the limits recently issued by European Pharmacopoeia 9.3 (2018 Sodium pertechnetate (99mTc) injection 4801-3). Moreover, the impact of 9xTc contaminant nuclides on the final image quality is thoroughly evaluated, analyzing the emitted high-energy gamma rays and their influence on the image quality. The spatial resolution of images from cyclotron-produced 99mTc acquired with a mini-gamma camera was determined and compared with that obtained using technetium-99m solutions eluted from standard 99Mo/99mTc generators. The effect of the increased image background contribution due to Compton-scattered higher-energy gamma rays (E γ > 200 keV), which could cause image-contrast deterioration, was also studied. It is concluded that, due to the high radionuclidic purity of cyclotron-produced 99mTc using 100Mo(p,2n)99mTc reaction at a proton beam energy in the range 15.7-19.4 MeV, the resulting image properties are well comparable with those from the generator-eluted 99mTc.

[((11))C]labelled radiopharmaceuticals as N-[(11)C]methyl-choline ([(11)C]choline), l-(S-methyl-[(11)C])methionine ([(11)C]methionine) and [(11)C]acetate have gained increasing importance in clinical PET and for the routine production of these radiopharmaceuticals, simple and reliable modules are needed to produce clinically relevant radioactivity. On the other hand, flexible devices are needed not only for the routine synthesis but also for more complex applications as the development of new tracers. The aim of this work was the adaptation of an Eckert Ziegler modular system for easy routine synthesis of [(11)C]choline, [(11)C]methionine and [(11)C]acetate using components that account for straightforward scaling up and upgrades.

In this paper we describe test methods useful to evaluate the performance of Ge/ Ga generators and to assess radionuclidic purity of the eluates. In our tests on the elution profile of the generator, we found that less than 8.0 ml are sufficient to recover > 98 % of the activity. Elution yield was conforming the manufacturer specifications, being initially > 85 % and still > 64 % after 7 months of use. 68Ge breakthrough in the eluates was limited; even if radionuclidic purity was > 99.99 % , further analyses, in addition to physical tests, is recommended in order to assess chemical impurities that could determine interferences in labelling procedures.

An experimental campaign was conducted to assess the neutron dose field inside the biological shield of the PETrace cyclotron at the S. Orsola-Malpighi hospital in Bologna. The results of the survey were analyzed to confirm that neutrons are emitted isotropically, thus validating the hypothesis of nuclear evaporation, as opposed to cascade, as the dominant mechanism for neutron emission.

In this paper, radionuclidic impurities generated during the bombardment of [18 O]water in the routine production of 2-[18F]fluoro-2-deoxy-d-glucose ([18F]FDG) were studied. In order to assess such impurities and the efficacy of purification methods through the different steps of the synthesis, samples of the target filters, purification columns, [18 O]water recovered after the synthesis, and the final solution was collected and their activities measured and analyzed by means of a gamma-ray spectrometry system. The data demonstrated that purification methods adopted for the synthesis provide the [18F]FDG radionuclidically pure, as requested by the EU Pharmacopeia.

Radionuclide activity meters ("dose calibrators") are ionization chambers designed to measure relatively high amount of activities which are normally contained in radiopharmaceuticals. However, in the current radiopharmacy practice, these radiation detectors have been proposed to be used in measurements of samples with lower activity, such as in routine quality control (QC) tests. To check the feasibility of such measurements, in this work we assessed the performance of four different devices in the lower range of detectability, by means of experimental measurements of a radioactive sample. Accuracy and precision of each device was evaluated as a function of the activity contained in the sample in order to estimate a threshold value, or minimum detectable activity (MDA), which, according to our operational definition, may be used to express the concept of Limit of Quantification (LoQ). Moreover, a generalized procedure for the estimation of the MDA was established, which, being device- and radionuclide-independent, it may be adopted by every laboratory. Our results showed a significant variability in the MDA achieved by different activity meters. Hence a single QC test may result feasible with one specific instrument, and not with another one. Moreover, feasibility depends also on the confidence level required for each test. For these reasons, each activity meter should be qualified for its MDA or LoQ by each laboratory according to a procedure such as that described in this paper.