Radioisotopes For Nuclear Medicine Research Articles

Abass Alavi: A giant in Nuclear Medicine turns 80 and is still going strong!

Abass Alavi: A giant in Nuclear Medicine turns 80 and is still going strong!

Manufacturing low enriched uranium metal by magnesiothermic reduction of UF4

This work presents an experimental description of thermal and physical studies to attain a practical manufacturing process of uranium metal enriched to 20% U235 (LEU – Low Enriched Uranium) by metallothermic reduction of UF4, with nuclear purity, for reduced amounts (1000g of uranium) and with radioactive safety. Uranium metal is needed to produce nuclear fuel elements based on uranium silicide (intermetallic U3Si2) and irradiation targets to produce Mo99. This process is a part of Brazilian efforts to fabricate the fuel elements for its research reactors, primarily aiming at the production of radioisotopes for nuclear medicine. The magnesiothermic reduction is influenced by variables which are related to the starting material UF4 and the thermal conditions for its reduction. These variables are investigated. The physical arrangement of the crucible/reduction reactor/furnace system and the management of the furnace thermal input in the reduction reactor during the heating were studied. Thermal simulation experiments provided delineation for the reactants’ thermal progress before the ignition of the metalothermic reaction. The heat input to the reduction system has proved to be the main variable that influenced the efficiency of the process. The levels of metallic yield and reproducibility have been improved, making the production process reproductive and economically viable. The typical yield in the production of uranium metal was above 80%. Unrecovered uranium metal is present in the MgF2 slag and can be recovered at the level of 96% yield. The process of recovering the uranium from the slag is also discussed.

GEANT4 simulation of cyclotron radioisotope production in a solid target

The use of radioisotopes in nuclear medicine is essential for diagnosing and treating cancer. The optimization of their production is a key factor in maximizing the production yield and minimizing the associated costs. An efficient approach to this problem is the use of Monte Carlo simulations prior to experimentation. By predicting isotopes yields, one can study the isotope of interest expected activity for different energy ranges. One can also study the target contamination with other radioisotopes, especially undesired radioisotopes of the wanted chemical element which are difficult to separate from the irradiated target and might result in increasing the dose when delivering the radiopharmaceutical product to the patient.The aim of this work is to build and validate a Monte Carlo simulation platform using the GEANT4 toolkit to model the solid target system of the South Australian Health and Medical Research Institute (SAHMRI) GE Healthcare PETtrace cyclotron. It includes a GEANT4 Graphical User Interface (GUI) where the user can modify simulation parameters such as the energy, shape and current of the proton beam, the target geometry and material, the foil geometry and material and the time of irradiation.The paper describes the simulation and presents a comparison of simulated and experimental/theoretical yields for various nuclear reactions on an enriched nickel 64 target using the GEANT4 physics model QGSP_BIC_AllHP, a model recently developed to evaluate with high precision the interaction of protons with energies below 200MeV available in Geant4 version 10.1. The simulation yield of the 64Ni(p,n)64Cu reaction was found to be 7.67±0.074 mCi·μA−1 for a target energy range of 9–12MeV. Szelecsenyi et al. (1993) gives a theoretical yield of 6.71mCi·μA−1 and an experimental yield of 6.38mCi·μA−1. The 64Ni(p,n)64Cu cross section obtained with the simulation was also verified against the yield predicted from the nuclear database TENDL and compared to experimental yield obtained from literature.

Cyclam Derivatives with a Bis(phosphinate) or a Phosphinato–Phosphonate Pendant Arm: Ligands for Fast and Efficient Copper(II) Complexation for Nuclear Medical Applications

Cyclam derivatives bearing one geminal bis(phosphinic acid), -CH2PO2HCH2PO2H2 (H2L(1)), or phosphinic-phosphonic acid, -CH2PO2HCH2PO3H2 (H3L(2)), pendant arm were synthesized and studied as potential copper(II) chelators for nuclear medical applications. The ligands showed good selectivity for copper(II) over zinc(II) and nickel(II) ions (log KCuL = 25.8 and 27.7 for H2L(1) and H3L(2), respectively). Kinetic study revealed an unusual three-step complex formation mechanism. The initial equilibrium step leads to out-of-cage complexes with Cu(2+) bound by the phosphorus-containing pendant arm. These species quickly rearrange to an in-cage complex with cyclam conformation II, which isomerizes to another in-cage complex with cyclam conformation I. The first in-cage complex is quantitatively formed in seconds (pH ≈5, 25 °C, Cu:L = 1:1, cM ≈ 1 mM). At pH >12, I isomers undergo nitrogen atom inversion, leading to III isomers; the structure of the III-[Cu(HL(2))] complex in the solid state was confirmed by X-ray diffraction analysis. In an alkaline solution, interconversion of the I and III isomers is mutual, leading to the same equilibrium isomeric mixture; such behavior has been observed here for the first time for copper(II) complexes of cyclam derivatives. Quantum-chemical calculations showed small energetic differences between the isomeric complexes of H3L(2) compared with analogous data for isomeric complexes of cyclam derivatives with one or two methylphosphonic acid pendant arm(s). Acid-assisted dissociation proved the kinetic inertness of the complexes. Preliminary radiolabeling of H2L(1) and H3L(2) with (64)Cu was fast and efficient, even at room temperature, giving specific activities of around 70 GBq of (64)Cu per 1 μmol of the ligand (pH 6.2, 10 min, ca. 90 equiv of the ligand). These specific activities were much higher than those of H3nota and H4dota complexes prepared under identical conditions. The rare combination of simple ligand synthesis, very fast copper(II) complex formation, high thermodynamic stability, kinetic inertness, efficient radiolabeling, and expected low bone tissue affinity makes such ligands suitably predisposed to serve as chelators of copper radioisotopes in nuclear medicine.

Evaluation the Synergistic Effect of High Dose Radiation of Radioiodine on the Immune System Suppressed By Cyclosporine

Cyclosporine is an immunosuppressant drug and iodine-131 is the diagnostic and therapeutic radioisotope in nuclear medicine. In this study, we aimed to evaluate the synergistic effect of high-dose radiation of radioiodine on the suppression of the immune system of BALB/cmice after receiving cyclosporine. A total of 50 mice BALB/c were randomly divided into two groups (control and test). Then, 50 mg kg-1 body weight cyclosporine was given to both groups. After 24 h, 0.5 mL iodine-131 was given to the test group and in the control group the same volume of normal saline was given. Data analysis was done using SPSS V.20 and p value <0.05 was considered significant. After 30 days of monitoring, only 2 deaths (8%) occurred in the control group, however 16 deaths (64%) were observed in the test group. Statistical analysis of the data indicated that there is a significant relationship (p value <0.0005) between injecting radioiodine and the mortality rate in mice receiving cyclosporine and proves the synergistic effect of these two factors. Based on the results, iodine-131 can be introduced as a stimulating and synergistic factor on the suppression of the immune system of BALB/cmice and subsequent death.

Comparison of DMSA scan 99 m and EC scan 99 m in diagnosis of cortical defect and differential renal function

Introduction:Diagnosis of renal cortical lesions by radioisotopes in nuclear medicine is one of the most common techniques and procedures can be performed by different radiotracer. However, all these materials are accurate in determining kidney function, but there are differences between them in the field. The purpose of this study was to evaluate the effectiveness of EC scans compared with DMSA scan in the detection of cortical lesions and DRF.Methods:65 cases, which have been referred for various reasons, for DMSA scans were enrolled. Patients 1 week after DMSA scan with the previous consent of the EC being scanned. The results were compared in terms of convergence as well as sensitivity, specificity, positive and negative predictive value of EC with respect to the results of DMSA scan.Results:PPV of EC was 100%, negative predictive value of EC was 68.75%, sensitivity of EC was 90.74% and specificity of EC was 100% in the detection of cortical lesions. DMSA scan and EC convergence rates result in cortical lesions in our study was high.Discussion:We suggest EC scan as an alternative to reduce the cost of therapy and radiation, but considering the benefits of DMSA scan, it could remain the gold standard method of diagnosis.

Bright perspectives for nuclear photonics

With the advent of new high-power, short-pulse laser facilities in combination with novel technologies for the production of highly brilliant, intense γ beams (like, e.g., Extreme Light Infrastructure – Nuclear Physics (ELI-NP) in Bucharest, MEGaRay in Livermore or a planned upgrade of the HIγS facility at Duke University), unprecedented perspectives will open up in the coming years for photonuclear physics both in basic sciences as in various fields of applications. Ultra-high sensitivity will be enabled by an envisaged increase of the γ-beam spectral density from the presently typical 102γ/eVs to about 104γ/eVs, thus enabling a new quality of nuclear photonics [1], assisted by new γ-optical elements [2]. Photonuclear reactions with highly brilliant γ beams will allow to produce radioisotopes for nuclear medicine with much higher specific activity and/or more economically than with conventional methods. This will open the door for completely new clinical applications of radioisotopes [3]. The isotopic, state-selective sensitivity of the well-established technique of nuclear resonance fluorescence (NRF) will be boosted by the drastically reduced energy bandwidth (<0.1%) of the novel γ beams. Together with a much higher intensity of these beams, this will pave the road towards a γ-beam based non-invasive tomography and microscopy, assisting the management of nuclear materials, such as radioactive waste management, the detection of nuclear fissile material in the recycling process or the detection of clandestine fissile materials. Moreover, also secondary sources like low-energy, pulsed, polarized neutron beams of high intensity and high brilliance [4] or a new type of positron source with significantly increased brilliance, for the first time fully polarized [5], can be realized and lead to new applications in solid state physics or material sciences.

Radiostrontium separation from sodium molybdate solution and its measurement using LSA: an application to radiopharmaceutical analysis

Technetium ( 99m Tc), a decay product of molybdenum ( 99 Mo), is employed as radioisotope in nuclear medicine. Several practical devices known as generators are commercially available which enable the user to separate the daughter from the parent radionuclide. The present study is focused on quality control of chro- matographic technetium generator. A properly constructed generator should comply with international requirements of radionuclide purity of 90 Sr/ 99 Mo B 6 9 10 -8 and 89 Sr/ 99 Mo B 6 9 10 -7 . For this purpose an analytical method was optimized to quantify radiostrontium ( 89 Sr and 90 Sr) in sodium molybdate (Na2 MoO4) solution, a fission product used for 99 Mo/ 99m Tc generators. Dowex 1 9 8 and alumina were used in sequence followed by tributyl phosphate extraction for radiostrontium separation. Cerenkov measurement of 89 Sr and 90 Sr (through its descendent 90 Y) was performed using Perkin Elmer Tricarb LSA 3170 with detection efficiency of 42 and 14 %, respectively. Since efficiency of Cerenkov counting is sensitive to presence of color, spectral index of sample was used to correct the counting efficiency. The chemical recovery for strontium was 22 % and for yttrium was 80 % as determined by inductively coupled plasma optical emission spectrometry. Lower limit of detection was found to be 6.3 and 14.4 Bq L -1 for 90 Sr and 89 Sr, respectively with 60 min counting time. Hence method can be applied successfully to analyze 89,90 Sr in fission molybdenum used as radiopharmaceutical with a relative error of \10 %.

Feasibility of a multi-purpose demonstration neutron source based on a compact superconducting spherical tokamak

Tokamak neutron sources would allow near term applications of fusion such as fusion–fission hybrid reactors, elimination of nuclear wastes, production of radio-isotopes for nuclear medicine, material testing and tritium production. The generation of neutrons with fusion plasmas does not require energetic efficiency; thus, nowadays tokamak technologies would be sufficient for such purposes. This paper presents some key technical details of a compact (∼1.8m3 of plasma) superconducting spherical tokamak neutron source (STNS), which aims to demonstrate the capabilities of such a device for the different possible applications already mentioned. The T-11 transport model was implemented in ASTRA for 1.5D simulations of heat and particle transport in the STNS core plasma. According to the model predictions, total neutron production rates of the order of ∼1015s−1 and ∼1013s−1 can be achieved with deuterium/tritium and deuterium/deuterium respectively, with 9MW of heating power, 1.4T of toroidal magnetic field and 1.5MA of plasma current. Engineering estimates indicate that such scenario could be maintained during ∼20s and repeated every ∼5min. The viability of most of tokamak neutron source applications could be demonstrated with a few of these cycles and around ∼100 cycles would be required in the worst cases.

Discharges of nuclear medicine radioisotopes in Spanish hospitals

Given the increasing use of radiopharmaceuticals in medicine, the aim of this paper is to determine radioactivity levels in the effluents of hospitals with Nuclear Medicine Departments. The radiological study of hospital discharges was carried out by gamma spectrometry, and liquid scintillation spectrometry to determine 14C and 3H contents. On March 9th and April 19th, 2010, daily radioactivity levels were monitored from 8:30 a.m. to 7:30 p.m. Each sample was collected at a specific control point of two major public hospitals in Granada (Spain). The analytical results show the presence of radionuclides such as 99mTc, 131I, 67Ga, and 111In.They are frequently used in nuclear medicine for diagnostic and/or therapeutic purposes. This study shows the differences between direct and after-storage discharges and also justifies the need of storage tanks in hospitals with nuclear medicine departments. Moreover, monitoring of 99mTc released at hospital control points can be a useful tool for optimizing the safety conditions of storage tanks and discharge of radionuclides.

Anti toxic effect of broccoli extract on stannous dichloride toxicity

Since Technetium-99m ((99m)Tc) has favorable physical and chemical characteristics, it is widely used radioisotope in Nuclear Medicine. However, stannous dichloride (SnCl(2)) has been widely used as a reducing agent in labeling procedure of pharmaceutical with radionuclide, it has been realized that SnCl(2) have genotoxic and cytotoxic effects on biological systems. In previous studies, it has been shown that some herbal extract can reduce genotoxic and cytotoxic effects of SnCl(2). In the present study, it is aimed to evaluate the effect of the broccoli extract on the survival of E. coli ATCC 25922 strain against to toxic effects of SnCl(2). Broccoli was extracted with methanol extraction. HPLC and TLC analysis of broccoli extract were performed. Then antitoxicity and dose response assays were performed on bacterial strain. The broccoli extract had dose dependent protective effect against SnCl(2) toxic effect on E. coli. The consumption of broccoli may alter the stannous dichloride toxicity. Broccoli extract may use as a new protective strategies against the toxic effect of SnCl(2) on patients who were taken (99m)Tc radiopharmaceuticals.

Production of medical radioisotopes with high specific activity in photonuclear reactions with γ-beams of high intensity and large brilliance

We study the production of radioisotopes for nuclear medicine in (γ,xn+yp) photonuclear reactions or (γ,γ′) photoexcitation reactions with high-flux [(1013–1015)γ/s], small diameter ∼(100 μm)2 and small bandwidth (ΔE/E≈10−3–10−4) γ beams produced by Compton back-scattering of laser light from relativistic brilliant electron beams. We compare them to (ion,xn+yp) reactions with (ion = p,d,α) from particle accelerators like cyclotrons and (n,γ) or (n,f) reactions from nuclear reactors. For photonuclear reactions with a narrow γ-beam the energy deposition in the target can be managed by using a stack of thin target foils or wires, hence avoiding direct stopping of the Compton and pair electrons (positrons). However, for ions with a strong atomic stopping only a fraction of less than 10−2 leads to nuclear reactions resulting in a target heating, which is at least 105 times larger per produced radioactive ion and often limits the achievable activity. In photonuclear reactions the well defined initial excitation energy of the compound nucleus leads to a small number of reaction channels and enables new combinations of target isotope and final radioisotope. The narrow bandwidth γ excitation may make use of the fine structure of the Pygmy Dipole Resonance (PDR) or fluctuations in γ-width leading to increased cross sections. Within a rather short period compared to the isotopic half-life, a target area of the order of (100 μm)2 can be highly transmuted, resulting in a very high specific activity. (γ,γ′) isomer production via specially selected γ cascades allows to produce high specific activity in multiple excitations, where no back-pumping of the isomer to the ground state occurs. We discuss in detail many specific radioisotopes for diagnostics and therapy applications. Photonuclear reactions with γ-beams allow to produce certain radioisotopes, e.g. 47Sc, 44Ti, 67Cu, 103Pd, 117m Sn, 169Er, 195m Pt or 225Ac, with higher specific activity and/or more economically than with classical methods. This will open the door for completely new clinical applications of radioisotopes. For example 195m Pt could be used to verify the patient’s response to chemotherapy with platinum compounds before a complete treatment is performed. Also innovative isotopes like 47Sc, 67Cu and 225Ac could be produced for the first time in sufficient quantities for large-scale application in targeted radionuclide therapy.

Optical observation of energy loss distribution and practical range of positrons from a water solution in a water‐equivalent phantom

The energy loss distribution of beta+ particles is closely related to their maximum penetration depth distribution and annihilation point distribution. The latter is of practical importance for positron emission tomography. Experimental data related to the energy loss distribution are important for comprehensive validation of physics and simulation models of beta+ interactions. In this paper the authors present a new experimental approach that allows them to visually observe the beta+ energy loss distribution of a solution of nuclear medicine radioisotopes in a plastic scintillator using an optical camera. The authors also report a set of the first experimental results. A water solution of 18F was localized in a small hole in a plastic scintillator (BC430). Optical imaging of the scintillator yielded visual images of the energy loss distribution with a submillimeter resolution. The radial dependence in the energy distribution was quantitatively measured by analysis of the images, and exponential fitting parameters were obtained. The authors observed that the results of Monte Carlo simulation with EGS5 (version 1.0.2) and GEANT4 (version 4.9.01.p01) were consistent with those obtained experimentally. The results of the Monte Carlo simulation indicated that for a linear scale, the energy loss distribution in the scintillator was approximately the same as that in water, and the relative shape of the energy loss distribution was close to those of the maximum penetration depth distribution and annihilation point distribution. This paper also presents discussions about the further possibilities of this optical imaging approach. Thus, optical observation of the beta+ energy loss distribution in a scintillator is a promising technique for visual and quantitative experimental studies of beta+ emission from a solution of radioisotopes that are used in nuclear medicine.

Fabrication of U&lt;sub&gt;3&lt;/sub&gt;Si&lt;sub&gt;2&lt;/sub&gt; Powder for Fuels Used in IEA-R1 Nuclear Research Reactor

Powdered uranium silicide (U3Si2) 20% U235 enriched is an intermetallic compound used as nuclear fuel material dispersed in aluminum to be the meat of fuel elements. U3Si2 powder is the state-of-the-art as nuclear fuel material mostly used in modern research reactors. Its recent established fabrication in IPEN replaced the previous ceramic powder U3O8 used in the fuel of IEAR1 (IPEN/CNEN, São Paulo, Brazil). The U3Si2 is a compound with 92.3%wtU and 7.7%wtSi. Its production is made by induction furnace melting using metallic uranium, produced by magnesiothermic reaction, and pure silicon. The induction furnace melts under argon controlled environment using zirconia crucible. Homogenization of liquid bath at 1800°C is a compromise between crucible resistance and homogenized melting, avoiding hazardous happenings. IPEN produced its first lot of enriched U3Si2 in September 2004, with a continuous fabrication ever since. This research work represents the ability of having fully Brazilian supply of this strategic and high cost nuclear material. The fuel quality meets the world quality standards required by International Atomic Energy Agency (IAEA) and RERTR standards. Brazilian production of U3Si2 powder not only closed the fuel cycle, from uranium mineral to fuel element, but also allowed higher productivity of nuclear medicine radioisotopes by IEA-R1.

Radiochemistry teaching and research activities in Brazil

Much concern has been expressed lately about the decline of teaching and research activities in radiochemistry in many countries, as was discussed in an IAEA Technical Meeting in Antalya, Turkey, in 2002, and also at MTAA-11 in Guildford, UK. In the IAEA meeting, a survey was presented about the current situation in different regions of the world (Eastern Europe, East and West Asia, Africa, North America and Latin America) by experts of each region. In the case of Brazil, which has nuclear research reactors and also cyclotrons in operation, the teaching and research activities in radiochemistry are concentrated in the three main institutes of the Brazilian Nuclear Energy Commission, in the University of Sao Paulo and in other universities, in different regions of the country. In the present paper, a closer look is given to the radiochemistry teaching and research activities that are being conducted nowadays in Brazil, comprising: number of radiochemistry courses and students being formed, main research areas being conducted, as well as research and production of radioisotopes for nuclear medicine, using nuclear reactors and cyclotrons.

Overview of University of Wisconsin Inertial-Electrostatic Confinement Fusion Research

In Inertial Electrostatic Confinement (IEC) devices, a voltage difference between concentric, nearly transparent spherical grids accelerates ions to fusion-relevant velocities. The University of Wisconsin (UW) operates two IEC devices: a cylindrical aluminum chamber and a spherical, water-cooled, stainless-steel chamber, with a power supply capable of 75 mA and 200 kV. The research program aims to generate fusion reaction products for various applications, including protons for creating radioisotopes for nuclear medicine and neutrons for detecting clandestine materials. Most IEC devices worldwide, including the UW devices, presently operate primarily in a pressure range (1-10 mtorr) that allows ions to make only a few passes through the core before they charge exchange and lose substantial energy or they collide with cathode grid wires. It is believed that fusion rates can be raised by operating at a pressure where neutral gas does not impede ion flow, and a helicon ion source has been developed to explore operation at pressures of ~0.05 mtorr. The UW IEC research group uses proton detectors, neutron detectors, residual gas analyzers, and spectroscopic diagnostics. New diagnostic techniques have also been developed, including eclipse disks to localize proton production and chordwires to estimate ion fluxes using power balance.

Production, processing and uses of 94mTc

AbstractTechnetium‐99m is a widely used radioisotope in nuclear medicine. The characteristics of this generator‐produced radionuclide (T1/2 = 6 h, Eγ = 140.5 keV, Iγ = 87.2 %) make it an ideal isotope for single photon emission computed tomography (SPECT). Technetium‐94m has a half‐life of 52 minutes and a positron branching ratio of 72 %, with Eβ+max of 2.44 MeV. Technetium‐94m undergoes the same chemistry as 99mTc, but has characteristics suitable for imaging using positron emission tomography (PET), with which more quantitative measurements of radiopharmaceutical uptake, biodistribution and clearance can be made.

Compilation of cross sections/thick target yields for 66Ga, 67Ga and 68Ga production using Zn targets up to 30 MeV proton energy

As a part of our continuing efforts to compile nuclear data relevant to the production of routinely used radioisotopes for nuclear medicine using small or medium size proton accelerators, we have reviewed the excitation functions and production data of 66Ga, 67Ga and 68Ga appearing in the literature up to the end of 1992. We have compiled the cross section data as well as calculated and measured thick target yields of 66Zn(p, n) 66Ga, 67Zn(p, 2n) 66Ga, 68Zn(p, 3n) 66Ga, 67Zn(p, n) 67Ga, 68Zn(p, 2n) 67Ga and 68Zn(p, n) 68Ga nuclear reactions in the energy range up to 30 MeV.

A remote device for de-encapsulating reactor-irradiated samples

A remotely operated motorized apparatus is described which facilitates the recovery of Ci levels of radioactive samples, such as 18F, 24Na, etc. without excessive personnel hand or whole-body dose, from sealed quartz ampoules which have been irradiated in a reactor in standard polythylene or aluminum containers. The device is operated in a small hot cell located in a walk-in hood and is used in the routine preparation of reactor-produced radioisotopes for nuclear medicine.

Technical Problems in the Use of Radioisotope in Nuclear Medicine

Technical Problems in the Use of Radioisotope in Nuclear Medicine