Radioisotopes For Nuclear Medicine Research Articles

Neurotensin (8-13) and Neuromedin N Neuropeptides Radiolabelling with Copper-64 Produced on Solid or Liquid Targets.

On the verge of a theranostic approach to personalised medicine, copper-64 is one of the emerging radioisotopes in nuclear medicine due to its exploitable nuclear and biochemical characteristics. The increased demand for copper-64 for preclinical and clinical studies has prompted the development of production routes. This research aims to compare the (p,n) reaction on nickel-64 solid versus liquid targets and evaluate the effectiveness of [64Cu]CuCl2 solutions prepared by the two routes. As new treatments for neurotensin receptor-overexpressing tumours have developed, copper-64 was used to radiolabel Neurotensin (8-13) and Neuromedin N. High-quality [64Cu]CuCl2 solutions were prepared using ACSI TR-19 and IBA Cyclone Kiube cyclotrons. The radiochemical purity after post-irradiation processing reached 99% (LT) and 99.99% (ST), respectively. The irradiation of a solid target with 11.8 MeV protons and 150 μAh led to 704 ± 84 MBq/μA (17.6 ± 2.1 GBq/batch at EOB). At the end of the purification process (1 h, 90.90% activity yield), the solution for peptide radiolabelling had a radioactive concentration of 1340.4 ± 70.1 MBq/mL (n.d.c.). The irradiation of a liquid target with 16.9 MeV protons and 230 μAh resulted in 3.7 ± 0.2 GBq/batch at EOB, which corresponds to an experimental production yield of 6.89 GBq.cm3/(g.µA)sat. Benefiting from a shorter purification process (40 min), the activity yielded 90.87%, while the radioactive concentration of the radiolabelling solution was lower (492 MBq/mL, n.d.c.). The [64Cu]CuCl2 solutions were successfully used for the radiolabelling of DOTA-NT(8-13) and DOTA-NN neuropeptides, resulting in a high RCP (>99%) and high molar activity (27.2 and 26.4 GBq/μmol for LT route compared to 45 and 52 GBq/μmol for ST route, respectively). The strong interaction between the [64Cu]Cu-DOTA-NT(8-13) and the colon cancerous cell lines HT29 and HCT116 proved that the specificity for NTR had not been altered, as shown by the uptake and retention data.

Study of the production of radioisotopes at IFMIF-DONES: 177Lu with deuterons

The production of radioisotopes for nuclear medicine is boosted by different international agencies. The International Fusion Materials Irradiation Facility - Demo Oriented NEutron Source (IFMIF-DONES) will be a infrastructure for testing under high neutron flux the materials to be used in future DEMO fusion reactor. IFMIF-DONES will generate neutrons by means of the impact of a high-power deuteron beam onto a lithium-jet target. There is an important effort to take advantage of the outstanding characteristics of the facility in terms of neutrons and deuterons. Thus, a complementary program is being developed, where the production of radioisotopes for medicine is one of the main applications. Here, we discuss the production of 177Lu with deuterons at IFMIF-DONES.

Production of 99Mo at IFMIF-DONES reusing the flux of neutrons

The accelerator-based neutron facility IFMIF-DONES is in the final design phase. At present, there is an important effort to optimize other secondary uses compatible with the main goal which is the validation of materials to be used in fusion reactors. One is the production of radioisotopes for nuclear medicine, in particular, Molybdenum-99. The decay of 99Mo produces Technetium-99 metastable. 99mTc is the most used radiopharmaceutical in nuclear medicine, in particular, it is used for SPECT imaging diagnostic. Currently, its production is compromised by the possible closure of some of the few nuclear reactors which produces it, and the search for new alternative ways to obtain it is urgent. This study has been carried out considering three types of molybdenum samples (natural and 95% enriched in 98Mo and 100Mo) and two different positions close to the neutron target. The results of the simulated production are promising and push further studies.

Simulation of the indirect reaction 92Mo(α, 2n)94Ru→m94Tc using TALYS and GEANT4 nuclear codes

Radioisotope 94mTc is a favorable positron emitter radioisotope in nuclear medicine that can be used for PET. Production of 94mTc via indirect reaction 92Mo(α, 2n)94Ru → 94mTc was predicted by SRIM, TALYS,and GEANT4 codes. SRIM and GEANT4 codes were employed to calculate the range of alpha particles into enriched 92Mo and to estimate the thickness of the target. Further, the GEANT4 and the TALYS codes were used to investigate the cross-section and production yield of the 92Mo(α, 2n)94Ru reaction. Four model methods were utilized to compute the cross-section by TALYS-1.95 code. The simulated results were compared with experimental values that reported by F.O. Denzler and found to be in good agreement between them. This observation indicates that the SRIM, the GEANT4 and the TALYS codes are suitable tools for prediction of producing radioisotopes via various reactions.

Optical fibre array detector to monitor irradiations for medical radioisotope production

Low-energy cyclotrons are in use worldwide to produce medical radioisotopes for nuclear medicine. Beam monitoring during the irradiation of targets is difficult due to the high-power density of low-energy protons, space limitations, and interference with beam delivery. Doped silica fibres are sensitive to prompt ionizing radiation from the bombarded target, and produce radiation induced luminescence (RIL) when exposed. The fibres can be attached to the outside of the target in a low-profile fibre array, ensuring efficient and safe operation. Here, we present the results from our prototype of such a fibre array. It consists of four Ce-doped fibres with a diameter of 200 μm and has been installed at the TR13 medical cyclotron at TRIUMF where it has been tested at different irradiation conditions.

Application of therapeutic linear accelerators for the production of radioisotopes used in nuclear medicine

Abstract This review paper summarizes the possibilities of the use of therapeutic linear electron accelerators for the production of radioisotopes for nuclear medicine. This work is based on our published results and the thematically similar papers by other authors, directly related to five medical radioisotopes as 99Mo/99mTc, 198Au, 186Re, 188Re, 117mSn, produced using therapeutic linacs. Our unpublished data relating to the issues discussed have also been used here. In the experiments, two types of radiation were included in the analysis of the radioisotope production process, i.e. the therapeutic twenty-megavolt (20 MV) X-rays generated by Varian linacs and neutron radiation contaminating the therapeutic beam. Thus, the debated radioisotopes are produced in the photonuclear reactions and in the neutron ones. Linear therapeutic accelerators do not allow the production of radioisotopes with high specific activities, but the massive targets can be used instead. Thus, the amount of the produced radioisotopes may be increased. Apart from linear accelerators, more and more often, the production of radioisotopes is carried out in small medical cyclotrons. More such cyclotrons are developed, built, and sold commercially than for scientific research. The radioisotopes produced with the use of therapeutic linacs or cyclotrons can be successfully applied in various laboratory tests and in research.

Neutronic Evaluation of Using a Thorium Sulfate Solution in an Aqueous Homogeneous Reactor

Radioisotope 99 Mo is one of the most essential radioisotopes in nuclear medicine. Its production in an Aqueous Homogeneous Reactor (AHR) could be potentially advantageous compared to the traditional technology, based on target irradiation in a heterogeneous reactor. An AHR conceptual design using low-enriched uranium for the production of 99 Mo has been studied in depth. So far, the possibility of replacing uranium with a non-uranium fuel, specifically a mixture of 232 Th and 233 U, has not been evaluated in the conceptual design. Therefore, the studies conducted in this article aim to evaluate the neutronic behavior of the AHR conceptual design using thorium sulfate solution. Here, the 232 Th- 233 U composition to guarantee ten years of operation without refueling, conversion ratio, medical isotopes production levels, and reactor kinetic parameters were evaluated, using the computational code MCNP6. It was obtained that 14 % 233 U enrichment guarantees the reactor operation for ten years without refueling. The conversion ratio was calculated at 0.14. The calculated 99 Mo production in the AHR conceptual design resulted in 24.4 % higher with uranium fuel than with thorium fuel.

Diagnostic and therapeutic radioisotopes in nuclear medicine: Determination of gamma-ray transmission factors and safety competencies of high-dense and transparent glassy shields

Abstract We present the findings of an extensive examination on newly designed CdO-rich and transparent glass shields for nuclear medicine facilities in lieu of traditional and unfavorable materials, such as lead and concrete. Gamma-ray transmission factors of newly designed glass shields are determined using a variety of diagnostic, therapeutic, and research radioisotopes, including 67Ga, 57Co, 111In, 201Tl, 99mTc, 51Cr, 131I, 58Co, 137Cs, 133Ba, and 60Co. A general-purpose Monte Carlo code MCNPX (version 2.7.0) is used to determine the attenuation parameters of different material thicknesses. Next, the findings are compared using a standard concrete shielding material. The results indicate that adding more CdO to the glass composition improves the overall gamma-ray attenuation properties. As a result, among the heavy and transparent glasses developed, the C40 sample containing 40% CdO exhibited the best gamma-ray absorption properties against all radioisotopes. Furthermore, the gamma-ray absorption characteristics of this created high-density glass were shown to be better to those of a standard and heavy concrete sample. It can be concluded that the newly developed CdO-rich and transparent glass sample may be used in medical radiation fields where the radioisotopes examined are used in daily clinical and research applications.

Radioisotope Products and the Medicine of the Future: an IAEA Perspective

The production and application of radioisotopes and radiopharmaceuticals is a major peaceful application of nuclear science and technology and has opened new gateways for nuclear medicine in critical human diseases. Advances in the production of radioisotopes via nuclear reactors, accelerators, fission, etc. has facilitated the access to these products for Member States. 99Mo, 131I, 177Lu are among the most essential radioisotopes for nuclear medicine and human health and their production and availability always has been an important theme for professional societies and the International Atomic Energy Agency (IAEA). In the meantime, a large list of theranostic radioisotopes including but not limited to 89Zr, 68Ga, 225Ac, Cu-series, Sc-series, Tb-series etc. has provided a powerful toolbox for clinicians and the IAEA is taking steps to ensure their safe and appropriate application in radiopharmacy. The Agency promotes the production and application routes, including research reactors, cyclotrons, linear accelerators, and other cutting-edge methods, according to international and national guidelines and regulations. The IAEA also conducts activities such as Coordinated Research Projects (CRPs), Technical Meetings (TMs), national/regional training courses and conferences, to support and join forces with international professional societies in the development of human resources and research and development activities. Development of databases and freely available publications for all Member States are other useful means to support Member States in radiopharmaceutical sciences.

Production of radioactive isotopes by photonnuclear reactions using the MT-22 microtrona

The wide circulation of application of radioisotopes in nuclear medicine, biochemical behavior of elements and environmental would be impossible without development of new methods for theirs of production. The FLNR MT-25 microtron is an important installation for the production of wide variety of man-made radioactive isotopes. On propose of academician of Flerov G.N. in 1980 year experiment were started at the microtron for producing of 123I.

Gamma-Ray Protection Properties of Bismuth-Silicate Glasses against Some Diagnostic Nuclear Medicine Radioisotopes: A Comprehensive Study.

This study aimed to perform an investigation for the potential implementation of bismuth silicate glasses as novel shield equipment instead of ordinary shields in nuclear medicine facilities. Accordingly, a group of Bi2O3 reinforced silicate glass system were investigated and compared with ordinary shields in terms of their gamma-ray attenuation properties in diagnostic nuclear medicine radioisotope energies emitted from 99mTc, 111In, 67Ga, 123I, 131I, 81mKr, 201Tl, 133Xe. Mass attenuation coefficient results for glass samples were calculated comparatively with the XCOM program and MCNPX code. The gamma-ray attenuation parameters such as half value layer (HVL), tenth value layer (TVL), mean free path (MFP), effective atomic number (Zeff) were obtained in the diagnostic gamma ray energy range from 75 to 336 keV. To confirm the attenuation performance of superior sample, obtained results were extensively compared with ordinary shielding materials. According to the results obtained, BISI6 glass sample with the highest Bi2O3 additive has an excellent gamma-ray protection.

Review: Production of nuclear medicine radioisotopes with ultra-intense lasers

In the last two decades, there has been a strong research interest in producing radioisotopes with ultra-intense lasers, as an application of laser-driven accelerators in nuclear medicine. Encouraging progress has been obtained in both experiments and simulations. This Review presents the results of several intense studied radioisotopes in detail, i.e., 18F, 11C, 13N, 15O, 99mTc, 64Cu, and 62Cu. As for other less studied radioisotopes, the results are summarized in Sec. II G. The results are listed in Tables I–VII along with laser intensities, maximum ion/photon energies, number of ions/photons per shot, reactions, and laser repetition rates and facilities. For research based on high repetition rate lasers, both single-shot and multi-shot productions are provided for the purpose of comparison. With key technologies implemented in new commissioning ultra-intense lasers, further experiments will definitely help moving this area forward, which will bring the realization of laser-driven radioisotope production closer.

Dosimetry study on Auger electron-emitting nuclear medicine radioisotopes in micrometer and nanometer scales using Geant4-DNA simulation

Purpose Dosimetry of Auger electron-emitting nuclear medicine radioisotopes on cellular and DNA scales is essential in order to assess the biological effects and damage of these radioisotopes to the DNA molecule. This study examined the effects of widely used radioisotopes in nuclear medicine and also two therapeutic radioisotopes in the scale of micrometer and nanometer. Methods In this paper, on cell scale (micrometer scale), the S-values for widely used Auger electron-emitting diagnostic radioisotopes 123I, 125I, 99mTc, 67Ga, 201Tl and 111In and two therapeutic radioisotopes of 131I and 211At in three different geometrical cell models (spherical, elliptical and cubic) were calculated using Geant4-DNA and the results were compared with the results of other simulation codes as well as the MIRD technique. On DNA scale (nanometer scale), the average number of DNA strand breaks (SSB and DSB) resulting from the direct and indirect effects was calculated for the specified radioisotope. Results The results showed that in the cell scale, S-values of the diagnostic radioisotopes were mostly greater than the S-values of the therapeutic radioisotope 131I, but they were less those of 211At. On DNA scale, two different geometric models of DNA molecule were simulated and the results of these two models were compared with each other, as well as with the literature. The results showed that the geometric shape of sugar-phosphate groups has a significant effect on the break rates of DNA molecules. Conclusions Among the widely used diagnostic radioisotopes, 201Tl and 125I had the greatest effect on the rate of SSBs and DSBs, respectively, while the 131I therapeutic radioisotope almost had no effect and therapeutic radioisotope of 211At had a moderate effect on the number of breaks.

An Overview of the Radioisotopes in Nuclear Medicine and Following Discussion on Promising Success of the Targeted Alpha Therapy

An Overview of the Radioisotopes in Nuclear Medicine and Following Discussion on Promising Success of the Targeted Alpha Therapy

Neutronic analysis of critical assembly for moly-99 production reactor based on mixed Th-U fuels

Neutronic analysis of nuclear reactor systems for 99Mo isotope production based on mixed thorium-uranium (Th-U) nitrate fuels has been done. The 99Mo isotope is a 99mTc generator which is the most widely used radioisotope in nuclear medicine. The proposed critical reactor system or critical assembly for 99Mo production or CAMOLYP is currently being studied in BATAN Yogyakarta as a continuation of the development of subcritical assembly for 99Mo production called SAMOP R&D program. The aim of this research is to obtain the optimum CAMOLYP core configurations, the core is an annular cylindrical provided by the outer ring. The method used is the calculation of the critical level of CAMOLYP for two reactor core configuration models, using the MCNPX computer code. The analysis result shows that there are various configurations of CAMOLYP core to achieve criticality condition. The best configuration of design-A is by using a Th-U nitrate density of 300 g Th-U/L or 120 g U/L, with a volume of 56.7057 L Th-U nitrate, surrounded by 38 TRIGA wt-12% in the outer ring (kef = 1.000334). While in Design-B, the best configuration is by using 24 fuel rods with a UN volume of 29.0978 L using the UN density of 300 g U/L (kef = 1.0562). The amount of 235U in the first core design is 3376.74 g, while the second design is 1724.046 g. Based on the total amount of 235U needed to achieve criticality, the author concluded that the best design for CAMOLYP is Design-B.

Effect of target density uncertainties on extracted experimental cross sections for the natTi(α, x)51Cr,46Sc reactions

The production optimization of medical radioisotopes for various applications requires careful analysis of all production parameters. The importance of such radioisotopes in nuclear medicine cannot be overemphasized. During cross section calculations, experimentalist sometimes find more than one reported literature value of a certain parameter, each leading to a different effect on the overall cross sections. In the present work, we considered the effect of slight variations of density values of titanium reported in the literature on the experimental cross sections. Several reported density values (4.50, 4.51 and 4.54 g/cm3) were found in the literature, but only the lowest and highest values (4.50 and 4.54 g/cm3) were considered in this study. The said densities were used to calculate the cross sections of Ti(a, x)51Cr,46Sc radioisotopes. The corresponding trends of the excitation functions of 51Cr and 46Sc were analyzed for the selected density values. The calculated cross sections from the different density values have been analyzed and compared graphically to show the level of variations, as well as comparison with theoretical TALYS code calculated cross sections. About 1 percent variation of the cross sections have been observed. Following the observed slight variation in cross sections effect of the various density values, we recommend that IAEA could update its nuclear data service database to include recommended density values for all metals using reliable sources.

Nanocrystalline titanium carbide/carbon composites as irradiation targets for isotopes production

Aiming at the preparation of porous targets of titanium carbide with excess of carbon, the sol-gel technique was used: Ti alkoxide precursors were reacted with a phenol-based resin under acid catalysed conditions of hydrolysis. The hydrolysis and condensation of titanium alkoxide provided titania nanoparticles covalently bonded with the phenolic resin behaving as carbon source, meanwhile preserving coordination with acetate ligands. This prepolymer containing Ti–O–Ti bonds, linkages between TiOx and the phenol resin and chelation with acetate groups were treated under inert atmosphere at high temperature (up to 1750 °C). The different stages of the carbothermal reduction were studied by FT-IR spectroscopy, X-ray diffraction spectroscopy and Scanning Electron Microscopy. Specific surface area and occurrence of meso/micro pores of the nanocomposites were determined by N2 physisorption, while the assessment of porosity was obtained by mass and volume measurements and He picnometry analysis. Results show that nano-sized TiC crystals embedded into a finely distributed carbon matrix were obtained, showing optimal functional properties for the potential application, such as optimized composition, high specific surface area and open porosity. The developed material can be used as a source of radioisotopes for nuclear medicine once irradiated in existing or future production facilities such as ISOLPHARM at INFN-LNL (Istituto Nazionale di Fisica Nucleare – Laboratori Nazionali di Legnaro).

Proceedings From the Clinical and Scientific Innovations in Oral and Maxillofacial Surgery Conference 2019

Proceedings From the Clinical and Scientific Innovations in Oral and Maxillofacial Surgery Conference 2019

Micro-dosimetry calculation of Auger-electron-emitting radionuclides mostly used in nuclear medicine using GEANT4-DNA

A large number of nuclear medicine radionuclides are Auger-electron-emitters and internal conversion electrons which can transmit significant doses to the patient during diagnosis. Therefore, the dosimetry of these radioisotopes is necessary for the evaluation of their biological effects and their use for treatment and targeted-radiotherapy.In this study, dosimetry calculation of a number of widely used radioisotopes in nuclear medicine was performed on a cellular scale using Geant4-DNA simulation.S-values of some of the diagnostic radioisotopes, including 123I, 125I, 99mTc, 67Ga, 201Tl, and 111In, were evaluated in a homogeneous spherical geometry model with unit density in which the cell and nucleus were concentric. The results revealed that S-values of these diagnostic radioisotopes were mainly greater than S-values of the radioisotope 131I, which emits β-particles; they were lower but can be compared with 211At (emitter of alpha particles) in the cellular scale. It shows better the importance of dosimetry calculation of diagnostic Auger-electron-emitting radioisotope in a cellular scale and their applicability in treatment. It should be noted that the S-values obtained out of the Geant4-DNA simulation are in line with the values of the other codes and the MIRD technique.

Status of Simulations for the Cyclotron Laboratory at the Institute for Nuclear Research and Nuclear Energy

The Institute for Nuclear Research and Nuclear Energy is preparing to operate a high-power cyclotron for production of radioisotopes for nuclear medicine, research in radiochemistry, radiobiology, nuclear physics, solid state physics. The cyclotron is a TR24 produced by ASCI, Canada, capable to deliver proton beams in the energy range of 15 to 24 MeV with current as high as 400 µA. Multiple extraction lines can be fed. The primary goal of the project is the production of PET and SPECT isotopes as 18F, 67,68Ga, 99mTc, etc.This contribution reports the status of the project. Design considerations for the cyclotron vault will be discussed for some of the target radioisotopes.