A new 124Xe irradiation system for 123I production.

Since 2001, Nuclear and Energy Research Institute IPEN-CNEN has produced weekly ultrapure iodine-123, using a manual irradiation system, fully developed in IPEN. Iodine-123 radiopharmaceuticals have been produced and distributed to hospitals and clinics of nuclear medicine, where several diagnostic imaging procedures for thyroid, brain and cardiovascular functions are performed. Due to the short half-life and emission of low-energy photons, this radioisotope becomes suitable for diagnosis in children. In the present work, the technical and constructive aspects of a new fully automated irradiation system, dedicated to 123I routine production, employing enriched xenon-124 gas as the target material is presented. This new system consists of a target, a water and helium cooling system, a cryogenic system, an electric power system, and a control and process monitoring unit, composed of supervisory software, connected to a programmable logic controller via personal computer. In this new concept, there is no need for human intervention during radioisotope production, reducing the possibility of eventual failures or incidents involving radioactive material. By using this new system, a specific yield of 2.70 mCi/µAh per irradiation was achieved in validation runs, and after three years of routine production of iodine-123, the system showed reliability and resilience.

First Results on Zinc Oxide Thick Film Deposition by Inverted Magnetron Sputtering for Cyclotron Solid Targets Production.

The magnetron sputtering technique has been investigated in recent years with ever-growing interest as a verifiable solid target manufacturing technology aimed at the production of medical radionuclides by using low-energy cyclotron accelerators. However, the possible loss of high-cost materials prevents access to work with isotopically enriched metals. The need for expensive materials for the supply of the growing demand for theranostic radionuclides makes the material-saving approach and recovery essential for the radiopharmaceutical field. To overcome the main magnetron sputtering drawback, an alternative configuration is proposed. In this work, an inverted magnetron prototype for the deposition of tens of µm film onto different substrates is developed. Such configuration for solid target manufacturing has been proposed for the first time. Two ZnO depositions (20-30 µm) onto Nb backing were carried out and analysed by SEM (Scanning Electron Microscopy) and XRD (X-ray Diffractogram). Their thermomechanical stability under the proton beam of a medical cyclotron was tested as well. A possible improvement of the prototype and the perspective of its utilisation were discussed.

Cyclotron production of 103Pd using a liquid target.

INTRODUCTION: In this work, we present the first feasibility study on the production of the medically important radionuclide 103Pd via the 103Rh(p,n)103Pd reaction by cyclotron irradiation of a liquid target. Using a liquid target removes the time consuming and complex dissolution process of rhodium post-irradiation due to its chemically inactive nature and thereby will improve the accessibility of this radioisotope. METHODS: Liquid targets made from Rh(NO3)3·×H2O salt dissolved in de-ionized water were irradiated using a 12 MeV beam at the TR13 cyclotron at TRIUMF, Vancouver. RESULTS: A maximum EOB activity of 1.03 ± 0.05 MBq was achieved with the tested conditions, sufficient for basic radiochemistry studies. An effective separation method using anion exchange chromatography is reported using 1 M HNO3 as an eluent for rhodium (90.1 ± 2.1 % recovery) and a 1:1 mixture of 0.5 M NH3 + NH4Cl palladium eluent (103.8 ± 2.3 % recovery). The solution showed good in-target pressure stability. However, the production efficiency decreased significantly with higher solution concentrations and irradiation lengths which puts into question the scaling potential of this method. CONCLUSION: This proof-of-concept study has demonstrated the potential for using liquid targets as complementary production method of 103Pd for research purposes. The liquid target route faces several scaling challenges but can nonetheless improve the availability of 103Pd and consequently aid in widening its utility for radiopharmaceuticals.

Design and assembly of an XY-type beam current monitor for cyclotron accelerators.

A water-cooled XY-type beam current monitor was designed, manufactured and assembled in a cyclotron accelerator beam transport line. Tests were performed, demonstrating that apparatus is an instrument of great assistance in proton beam position. The XY-type beam current monitor has been widely used in liquid target irradiations, employing irradiation system which were originally designed for irradiations on 18 MeV cyclotron accelerator (Cyclone 18, IBA) only, however, with this apparatus, the target may be exchanged between the 30 MeV (Cyclone 30, IBA) and 18 MeV cyclotrons.

Current State of 44Ti/44Sc Radionuclide Generator Systems and Separation Chemistry.

In recent years, there has been an increased interest in 44Ti/44Sc generators as an onsite source of 44Sc for medical applications without needing a proximal cyclotron. The relatively short half-life (3.97 hours) and high positron branching ratio (94.3%) of 44Sc make it a viable candidate for positron emission tomography (PET) imaging. This review discusses current 44Ti/44Sc generator designs, focusing on their chemistry, drawbacks, post-elution processing, and relevant preclinical studies of the 44Sc for potential PET radiopharmaceuticals.

Exploration of producing Uranium-232 for use as a tracer in uranium fuels.

This study examines the production of 232U via neutron irradiation. Uranium-232 is considered for use as a tracer in nuclear fuel. However, a source of 232U is needed. This study examines the production of 232U via neutron irradiation of targets constructed out of either 231Pa or 230Th in the High Flux Isotope Reactor. HFIR targets were modeled in MCNP to determine 232U yield in protactinium and thorium targets. Flux tallies were used inside the targets to determine the neutron flux inside each target. This flux was then used in SCALE 6.2 ORIGEN to determine the 232U yield, as well as the buildup of the byproduct 233U. Several 230Th enrichments were examined to determine how 232U yield is affected by 230Th enrichment, as well as the effects of the presence of 232Th. The buildup of 228Th and 229Th in thorium targets was also examined, as these isotopes may impact the feasibility of recycling of thorium target materials.

A feasibility study of the therapeutic application of a mixture of 67/64 Cu radioisotopes produced by cyclotrons with proton irradiation.

PURPOSE: 64 Cu and 67 Cu radioisotopes have nuclear characteristics suitable for nuclear medicine applications. The production of 64 Cu is already well established. However, the production of 67 Cu in quantities suitable to conduct clinical trials is more challenging as it leads to the coproduction of other Cu isotopes, in particular 64 Cu. The aim of this study is to investigate the possibility of using a CuCl2 solution with a mixture of 67/64 Cu radioisotopes for therapeutic purposes, providing an alternative solution for the cyclotron production problem. METHODS: Copper radioisotopes activities were calculated by considering proton beam irradiation of the following targets: (i) 70 Zn in the energy range 70-45 MeV; (ii) 68 Zn in the energy range 70-35 MeV; (iii) a combination of 70 Zn (70-55 MeV) and 68 Zn (55-35 MeV). The contribution of each copper radioisotope to the human-absorbed dose was estimated with OLINDA/EXM software using the biokinetic model for CuCl2 published by ICRP 53. The total absorbed dose generated by the 67/64 CuCl2 mixture, obtained through different production routes, was calculated at different times after the end of the bombardment (EOB). A simple spherical model was used to simulate tumors of different sizes containing uniformly distributed 67/64 Cu mixture and to calculate the absorbed dose of self-irradiation. The biological damage produced by 67 Cu and 64 Cu was also evaluated through cellular dosimetry and cell surviving fraction assessment using the MIRDcell code, considering two prostate cancer cell lines with different radiosensitivity. RESULTS: The absorbed dose to healthy organs and the effective dose (ED) per unit of administered activity of 67 CuCl2 are higher than those of 64 CuCl2 . Absorbed dose values per unit of administered activity of 67/64 CuCl2 mixture increase with time after the EOB because the amount of 67 Cu in the mixture increases. Survival data showed that the biological damage caused per each decay of 67 Cu is greater than that of 64 Cu, assuming that radionuclides remain accumulated in the cell cytoplasm. Sphere model calculations demonstrated that 64 Cu administered activity must be about five times higher than that of 67 Cu to obtain the same absorbed dose for tumor mass between 0.01 and 10 g and about 10 times higher for very small spheres. Consequently, the 64 CuCl2 -absorbed dose to healthy organs will reach higher values than those of 67 CuCl2 . The supplemental activity of the 67/64 CuCl2 mixture, required to get the same tumor-absorbed dose produced by 67 CuCl2 , triggers a dose increment (DI) in healthy organs. The waiting time post-EOB necessary to keep this DI below 10% (t10% ) depends on the irradiation methods employed for the production of the 67/64 CuCl2 mixture. CONCLUSIONS: A mixture of cyclotron produced 67/64 Cu radioisotopes proved to be an alternative solution for the therapeutic use of CuCl2 with minimal DI to healthy organs compared with pure 67 Cu. Irradiation of a 70 Zn+68 Zn target in the 70-35 MeV proton energy range for 185 h appears to be the best option from among all the production routes investigated, as it gives the maximum amount of activity, the shortest t10% (10 h), and less than 1% of 61 Cu and 60 Cu impurities.

Measurement of the 160Gd(p,n)160Tb excitation function from 4-18 MeV using stacked-target activation.

The  160Gd(p,n)160Tb excitation function was measured between 4-18 MeV using stacked-target activation at Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. Nine copper and eight titanium foils served as proton fluence monitor foils, using the  natCu(p,x)65Zn,  natTi(p,x)48V, and  natTi(p,x)46Sc monitor standards, respectively. Variance minimization using an MCNP v.6.2 model reduced the systematic uncertainties in proton energy and fluence. A priori predictions of the  160Gd(p,n) reaction using ALICE, CoH, EMPIRE, and TALYS, as well as the TENDL database, are compared to the experimentally measured values.

Accelerator Production of Scandium Radioisotopes: Sc-43, Sc-44, and Sc-47.

Scandium radioisotopes are increasingly considered viable radiolabels for targeted molecular imaging (Sc-43, Sc-44) and therapy (Sc-47). Significant technological advances have increased the quantity and quality of available radioscandium in the past decade, motivated in part by the chemical similarity of scandium to therapeutic radionuclides like Lu-177. The production and radiochemical isolation techniques applied to scandium radioisotopes are reviewed, focusing on charged particle and electron linac initiated reactions and using calcium and titanium as starting materials.

Initial phase of Pu-238 production in Idaho National Laboratory.

The initial phase of Plutonium-238 (Pu-238) production for radioisotope thermal generation is described here in detail. Two dosimeters with/without cadmium sleeve containing Neptunium-237 (Np-237) and other neutron sensors were inserted into the low power Advanced Test Reactor Critical (ATRC) for Pu-238 production testing. The gamma-ray energy measurements from Np-238, the short-lived intermediate product, confirmed that sizable amount of Pu-238 can be produced in the full power Advanced Test Reactor (ATR). The Pu-238 production determined from the irradiation experiment was in accord with the modelling predictions. Detailed studies of the Au/Cu sensors for thermal and epithermal flux analysis and their consistency for sensors in different locations, bare and in Cd-sleeve, provided confidence in the Pu-238 production data.

Cross section of the 96Zr(α,n)99Mo reaction induced by α-particles beams on natZr targets.

The excitation function of the 96Zr (α,n)99Mo reaction was determined using the stacked-foil activation technique. For the experiments, two stacks with metal foils of Cu, Ti and Zr of natural isotopic composition were irradiated independently with a 27.2 MeV α-particle beam. The characteristics of the primary beam and its verification along each stack were determined according to the well-known natCu(α,x)65Zn, natCu(α,x)66,67Ga, and natTi (α,x)51Cr monitor reactions. It was deduced that the expected production yield from 99Mo by irradiating 96Zr targets with a 23.8 MeV alpha particle beam for 1 h is 1.77 MBq/µA. According to the results, irradiation characteristics are proposed to produce 99Mo with high specific activity.

BNCT research activities at the Granada group and the project NeMeSis: Neutrons for medicine and sciences, towards an accelerator-based facility for new BNCT therapies, medical isotope production and other scientific neutron applications.

The Granada group in BNCT research is currently performing studies on: nuclear and radiobiological data for BNCT, new boron compounds and a new design for a neutron source for BNCT and other applications, including the production of medical radioisotopes. All these activities are described in this report.

Production of radiometals in liquid targets.

Over the last several years, the use of radiometals has gained increasing relevance in supporting the continuous development of new, complementary and more specific biological targeting agents. Radiopharmaceuticals labelled with radiometals from elements such as Tc, Zr, Y, Ga and Cu received increasing attention as they find application in both diagnostic SPECT and PET imaging techniques and radiotherapeutic purposes. Such interest stems from the wide variety of radionuclides available with distinct and complementary nuclear decay characteristics to choose from with unequalled specificity, but can also be explained by growing demand in targeted radionuclide therapy. As a result, as routine supply of these radiometals becomes mandatory, studies describing their production processes have expanded rapidly. Although most radiometals are traditionally provided by the irradiation of solid targets in specialized cyclotrons, recently developed techniques for producing radiometals through the irradiation of liquid targets have received growing attention due to compatibility with commonly available small medical cyclotrons, promising characteristics and encouraging results. Irradiating liquid targets to produce radiometals appears as a fast, reliable, convenient and cost-efficient alternative to the conventional solid target techniques, characterized by complex and time-consuming pre- and post-irradiation target handling. Production of radiometals in liquid targets incorporated to complete manufacturing processes for daily routine is already recognized as a viable alternative and complementary supply methodology to existing solid target based infrastructures to satisfy growing clinical demands. For instance, several sites already use the approach to produce 68Ga-radiopharmaceuticals for clinical use. This review article covers the production of common radiometals with clinical potential through the irradiation liquid targets. A comparison with the traditional solid target irradiation methods is presented when relevant.

[Work at the cyclotron for radioisotope production: exposure risks and prevention measures]. / Il lavoro al ciclotrone per produzione di radioisotopi: rischi espositivi e misure di prevenzione.

SUMMARY: In this article, there will be presented the main aspects of the radiation protection for a site dedicated to the production of radioisotopes by cyclotron for medical use. After analyzing the design parameters of the site for the operation from a point of view of the radiation protection of the operators, the population and the environment (shielding around the accelerator, ventilation system, chimney releases etc.), we will focus our attention to the handling of unsealed sources produced, with particular regard to the operating procedures. The final part will be dedicated to intervention in exceptional operating conditions (entring the bunker due to a machine failure, breaking of a vial during transfer from the hot cell to the transport container and accidental contamination by an operator).

Pre-feasibility Study for Establishing Radioisotope and Radiopharmaceutical Production Facilities in Developing Countries.

BACKGROUND: A significant number of developing countries have no facilities to produce medical radioisotopes and radiopharmaceuticals. OBJECTIVE: In this paper we show that access to life-saving radioisotopes and radiopharmaceuticals and the geographical distribution of corresponding infrastructure is highly unbalanced worldwide. METHODS: We discuss the main issues which need to be addressed in order to establish the production of radioisotopes and radiopharmaceuticals, which are especially important for developing countries as newcomers in the field. The data was gathered from several sources, including databases maintained by the International Atomic Energy Agency (IAEA), World Health Organization (WHO), and other international organizations; personal interactions with representatives in the nuclear medicine field from different regions of the world; and relevant literature. RESULTS: Developing radioisotope and radiopharmaceutical production program and installing corresponding infrastructure requires significant investments, both man-power and financial. Support already exists to help developing countries establish their medical radioisotope production installations from several organizations, such as IAEA. CONCLUSION: This work clearly shows that access to life-saving radioisotopes and the geographical distribution of corresponding infrastructure is highly unbalanced. Technology transfer is important as it not only immediately benefits patients, but also provides employment, economic activity and general prosperity in the region to where the technology transfer is implemented.

Calculation of productions of medical 201Pb, 198Au, 186Re, 111Ag, 103Pd, 90Y, 89Sr, 77Kr, 77As, 67Cu, 64Cu, 47Sc and 32P nuclei used in cancer therapy via phenomenological and microscopic level density models.

In the present study, we have widely investigated the production of nuclei used in cancer therapy for both phenomenological and microscopic level density models via TALYS and EMPIRE codes. To estimate the production of the radioisotope, we calculated the cross-section curves of the reaction and the integral yield curves for nine level density models using the cross-sections and the mass-stopping powers acquired from X-PMSP program in the particle beam current of 1 µA and irradiation time 1 h. To discuss the obtained results on the basis of the cross-sections and the integral yields curves, the results were compared with the experimental data and the recommended data in the literature.

Production of Sc medical radioisotopes with proton and deuteron beams.

Proton and deuteron beams (15.3 and 6.8 MeV, respectively) extracted from the PETtrace medical cyclotron at the Radiopharmaceuticals Production and Research Centre in the University of Warsaw, Heavy Ion Laboratory, 28 MeV protons from the C30 cyclotron at the National Centre for Nuclear Research, Swierk, near Warsaw and 33 MeV protons from the ARRONAX accelerator, Nantes were used to produce and investigate the medically interesting Sc radioisotopes. Both natural and isotopically enriched CaCO3 and TiO2 targets were used (42Ca, 43Ca, 44Ca, 48Ca, 48Ti). The production efficiency and isotopic purity were determined and are reported here for the highest commercially available enrichments of the target material. The Thick Target Yield, Activities at the End of Bombardment (EOB) and the relative activities of produced impurities at EOB are reported for 43Sc, 44gSc, 44mSc and 47Sc produced with particle energies below 33 MeV.

Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor.

The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors (HTGRs) was investigated. A high-temperature engineering test reactor (HTTR), which is located in Japan at Oarai-machi Research and Development Center, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the 125I production could be maximized and the 126I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 1.8 × 105GBq/y of 125I production.

The feasibility study of 177Lu production in Miniature Neutron Source Reactors using a multi-stage approach in Isfahan, Iran.

Miniature neutron source reactors (MNSRs) are among the safest and economic research reactors with potentials to be used for neutron studies. This manuscript explores the feasibility of 177Lu production in Isfahan MNSR reactor using direct production route. In this study, to assess the specific activity of the produced radioisotope, a simulation was carried out through the MCNPX2.6 code. The simulation was validated by irradiating a lutetium disc-like (99.98 chemical purity) at the thermal neutron flux of 5 × 1011 ncm2s-1 and an irradiation time of 4min. After the spectrometry of the irradiated sample, the experimental results of 177Lu production were compared with the simulation results. In addition, factor from the simulation was extracted by replacing it in the related equations in order to calculate specific activity through a multi-stage approach, and by using different irradiation techniques. The results showed that the simulation technique designed in this study is in agreement with the experimental approach (with a difference of approximately 3%). It was also found that the maximum 177Lu production at the maximum flux and irradiation time allows access to 723.5mCi/g after 27 cycles. Furthermore, the comparison of irradiation techniques showed that increasing the irradiation time is more effective in 177Lu production efficiency than increasing the number of irradiation cycles. In a way that increasing the irradiation time would postpone the saturation of the productions. On the other hand, it was shown that the choice of an appropriate irradiation technique for 177Lu production can be economically important in term of the effective fuel consumption in the reactor.

47Sc production development by cyclotron irradiation of 48Ca.

The therapeutic radionuclide 47Sc was produced through the 48Ca(p,2n) channel on a proton beam accelerator. The obtained results show that the optimum proton energies are in the range of 24-17 MeV, giving the possibility to produce 47Sc radionuclide containing 7.4% of 48Sc. After activation, the powdery CaCO3 target material was dissolved in HCl and scandium isotopes were isolated from the targets. The performed separation experiments indicate that, due to the simplicity of the operations and the chemical purity of the obtained 47Sc the best separation process is when scandium radioisotopes are separated on the 0.2 µm filter.