99 Mo recoil atoms yield in the 100 Mo(p,x) reaction on cyclotron irradiation of Mo nanofilms.

We study the possibility of increasing the specific activity of 99 Mo produced by irradiating molybdenum targets with the use of the Szilard-Chalmers effect. According to the Szilard-Chalmers effect, recoil atoms of the 99 Mo radionuclide can be produced in nuclear reactions and retained in surrounding buffer substance. The objective of our work is to measure the yield of recoil atoms in the buffer as a function of the molybdenum layer thickness. The yield of recoil 99 Mo atoms in the 100 Mo(p,x)99 Mo nuclear reaction as a function of metallic molybdenum nanolayer thickness has been measured. Measurements were carried out after irradiation of nanolayers in the U-150 cyclotron with 28-MeV protons. Nanolayers of molybdenum with thickness of 30 to 220 nm were produced by magnetron sputtering on sapphire plates. The yield of the mass of 99 Mo recoil atoms was found to be the highest with the molybdenum layer thickness of 80 ± 5 nm. The enrichment factor (EF) has been defined as a ratio of the 99 Mo collector specific activity (SA) to the entire target SA before separation as EF = 1200 ± 150. The specific activity of 99 Mo produced in the collector was estimated to be approximately equal to 25 Ci·g-1 .

Neutronic simulation of medical radioisotope 99Mo and 177Lu production in IPR 14 MeV neutron generator facility.

An accelerator based 14 MeV neutron generator is commissioned at Institute for Plasma Research India. The generator is based on the linear accelerator concept where the deuterium ion beam impinged to the tritium target to produce neutrons. The generator is designed to produce 1 × 1012 neutrons per sec. The 14 MeV neutron source facilities are an emerging tool for the lab scale experiments and research. In order to utilize the generator for the welfare of humanity, the assessment is made for the production of medical radioisotopes using the neutron facility. The usage of radioisotopes in the treatment and diagnosis of a disease is an important factor in the healthcare sector. A series of calculations are conducted to generate radioisotopes, especially 99Mo and 177Lu those are having huge applications in the medical and pharmaceutical industries. 99Mo can be also generated through neutron reactions 98Mo(n, g)99Mo and 100Mo(n, 2n)99Mo apart from fission reaction. The cross section of 98Mo(n, g)99Mo is high in the thermal energy range whereas 100Mo(n,2n)99Mo occurs at a high energy range. 177Lu can be produced using the reactions 176Lu (n, g)177Lu and 176Yb (n, g)177Yb. The cross section of both 177Lu production routes is higher at thermal energy range. The neutron flux level near the target is around 1010 cm-2s-1. In order to enhance production capabilities, the neutron energy spectrum moderators are used to thermalize the neutrons. The materials used as a moderator are beryllium, HDPE, graphite, etc. Moderators enhance the capabilities of medical isotope production in neutron generators.

Green synthesis of alumina nanoparticle using Hibiscus rosa-sinensis leaf extract as a candidate for molybdenum-99 adsorbent.

Al2O3 nanoparticle is effectively used as an adsorbent for the low specific activity of molybdenum-99 (99Mo). The Al2O3 nanoparticle was synthesized by the green synthesis method using Hibiscus rosa-sinensis leaf extract (HRE). The Al2O3 nanoparticle synthesized using 10% of the HRE has a crystallite size of 4.9 nm, a surface area of 254.6 m2/g, a pore size of 9.1 nm, a pore volume of 0.58 cm3/g and has a Mo adsorption capacity of 43.4 ± 6.1 mg Mo/g.

A quantitative description of the compatibility of technetium-selective chromatographic technetium-99m separation with low specific activity molybdenum-99.

Technetium-99m generators employing a technetium-selective stationary phase are a chromatographic instrument developed for use with 99Mo having low specific activity (LSA); particularly, 99Mo produced by electron accelerators. This paper presents a mathematical description of technetium-selective chromatographic (TSC) 99mTc separation and analyzes its compatibility with LSA 99Mo. We developed a theoretical formula for TSC 99mTc separation by discretizing its pertechnetate selectivity, and validated it using an electron linear accelerator and activated carbon-based TSC (AC-TSC) 99mTc generators. We confirmed that the activity concentration of 99mTc obtained from a TSC 99mTc generator can be calculated directly from its input 99Mo activity regardless of the 99Mo specific activity. The formula corroborates that TSC 99mTc separation is compatible with LSA 99Mo, and has a practical application in estimating the number of TSC 99mTc generators required for 99mTc demand of interest.

Future development of global molybdenum-99 production and saving of atmospheric radioxenon emissions by using nuclear fusion-based approaches.

Technetium-99m, the decay product of molybdenum-99, is the most used medical isotope in diagnostic imaging. The future disruptions of molybdenum-99 supply, due to the final shut down of some old producing reactors, has led some current global supplies to plan the expansion of their production capacity. While other countries are developing own production facilities to supply their domestic demand. The global increase of molybdenum-99 production in the coming years could increase by about five times the current demand, with about the 50 percent of additional production in North America. Xenon radionuclides are an inevitable by-product of the nuclear plants production, and their periodically release into the atmosphere, contribute to the background that is also revealed by the IMS stations of the CBTO treaty. In this framework, the development of new technologies, posing no risk in relation to nuclear proliferation and do not result in emissions of radioxenon, could mitigate the issues related to the forecast increase of molybdenum-99 production worldwide. In Italy, an alternative 99Mo production project, the project ENEA Sorgentina, based on the irradiation of molybdenum by neutrons produced by a deuterium-tritium nuclear fusion process, is under development. This facility will not release radioxenon into the atmosphere, so it will not affect the background value in the atmosphere in Southern Europe.

Fusion-Based Neutron Generator Production of Tc-99m and Tc-101: A Prospective Avenue to Technetium Theranostics.

Presented are the results of 99mTc and 101Tc production via neutron irradiation of natural isotopic molybdenum (Mo) with epithermal/resonance neutrons. Neutrons were produced using a deuterium-deuterium (D-D) neutron generator with an output of 2 × 1010 n/s. The separation of Tc from an irradiated source of bulk, low-specific activity (LSA) Mo on activated carbon (AC) was demonstrated. The yields of 99mTc and 101Tc, together with their potential use in medical single-photon emission computed tomography (SPECT) procedures, have been evaluated from the perspective of commercial production, with a patient dose consisting of 740 MBq (20 mCi) of 99mTc. The number of neutron generators to meet the annual 40,000,000 world-wide procedures is estimated for each imaging modality: 99mTc versus 101Tc, D-D versus deuterium-tritium (D-T) neutron generator system outputs, and whether or not natural molybdenum or enriched targets are used for production. The financial implications for neutron generator production of these isotopes is also presented. The use of 101Tc as a diagnostic, therapeutic, and/or theranostic isotope for use in medical applications is proposed and compared to known commercial nuclear diagnostic and therapeutic isotopes.

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.

Production cross sections of Mo, Nb and Zr radioisotopes from α-induced reaction on natZr.

Cross sections of α-induced reactions on natural zirconium were measured up to 50 MeV using the stacked-foil technique, activation method and high resolution γ-ray spectrometry. The production cross sections of 93m,99Mo, 90g,92m,95g,95m,96Nb and 88,89g,95Zr were determined and compared with other experimental data measured earlier and result of theoretical calculations. The integral thick target yield of 99Mo was deduced from the measured cross section data.

Development of silver nanoparticle-doped adsorbents for the separation and recovery of radioactive iodine from alkaline solutions.

Removing radioactive iodine from solutions containing fission products is essential for nuclear facility decontamination, radioactive waste treatment, and medical isotope production. For example, the production of high-purity fission 99Mo by irradiation of 235U with neutrons involves the removal of iodine from an alkaline solution of the irradiated target (which contains numerous fission products and a large quantity of aluminate ions) using silver-based materials or anion-exchange resins. To be practically applicable, the utilized iodine adsorbent should exhibit a decontamination factor of at least 200. Herein, the separation of radioactive iodine from alkaline solutions was achieved using alumina doped with silver nanoparticles (Ag NPs). Ag NPs have a larger surface area than Ag powder/wires and can thus adsorb iodine more effectively and economically, whereas alumina is a suitable inert support that does not adsorb 99Mo and is stable under basic conditions. The developed adsorbents with less impurities achieved iodine removal and recovery efficiencies of 99.7 and 62%, respectively, thus being useful for the production of 131I, a useful medical isotope.

Application of computational models to estimate organ radiation dose in rainbow trout from uptake of molybdenum-99 with comparison to iodine-131.

This study compares three anatomical phantoms for rainbow trout (Oncorhynchus mykiss) for the purpose of estimating organ radiation dose and dose rates from molybdenum-99 ((99)Mo) uptake in the liver and GI tract. Model comparison and refinement is important to the process of determining accurate doses and dose rates to the whole body and the various organs. Accurate and consistent dosimetry is crucial to the determination of appropriate dose-effect relationships for use in environmental risk assessment. The computational phantoms considered are (1) a geometrically defined model employing anatomically relevant organ size and location, (2) voxel reconstruction of internal anatomy obtained from CT imaging, and (3) a new model utilizing NURBS surfaces to refine the model in (2). Dose Conversion Factors (DCFs) for whole body as well as selected organs of O. mykiss were computed using Monte Carlo modeling and combined with empirical models for predicting activity concentration to estimate dose rates and ultimately determine cumulative radiation dose (µGy) to selected organs after several half-lives of (99)Mo. The computational models provided similar results, especially for organs that were both the source and target of radiation (less than 30% difference between all models). Values in the empirical model as well as the 14 day cumulative organ doses determined from (99)Mo uptake are compared to similar models developed previously for (131)I. Finally, consideration is given to treating the GI tract as a solid organ compared to partitioning it into gut contents and GI wall, which resulted in an order of magnitude difference in estimated dose for most organs.

GEANT 4 simulation of (99)Mo photonuclear production in nanoparticles.

GEANT 4 Monte-Carlo simulation toolkit is used to study the kinematic recoil method of (99)Mo photonuclear production. Simulation for bremsstrahlung photon spectrum with maximum photon energy 30MeV showed that for MoO3 nanoparticle escape fraction decreases from 0.24 to 0.08 when nanoparticle size increases from 20nm to 80nm. For the natural molybdenum and pure (100)Mo we obtained the lower values: from 0.17 to 0.05. The generation of accompanying molybdenum nuclei is significantly lower for pure (100)Mo and is about 3.6 nuclei per single (99)Mo nucleus, while natural molybdenum nanoparticle produce about 48 accompanying nuclei. Also, we have shown that for high-energy photons escape fraction of (99)Mo decreases, while production of unwanted molybdenum isotopes is significantly higher.

A novel monolithic LEU foil target based on a PVD manufacturing process for 99Mo production via fission.

99Mo is the most widely used radioactive isotope in nuclear medicine. Its main production route is the fission of uranium. A major challenge for a reliable supply is the conversion from highly enriched uranium (HEU) to low enriched uranium (LEU). A promising candidate to realize this conversion is the cylindrical LEU irradiation target. The target consists of a uranium foil encapsulated between two coaxial aluminum cladding cylinders. This target allows a separate processing of the irradiated uranium foil and the cladding when recovering the 99Mo. Thereby, both the costs and the volume of highly radioactive liquid waste are significantly reduced compared to conventional targets. The presented manufacturing process is based on the direct coating of the uranium on the inside of the outer cladding cylinder. This process was realized by a cylindrical magnetron enhanced physical vapor deposition (PVD) technique. The method features a highly automated process, a good quality of the resulting uranium foils and a high material utilization.

Effects study on the thermal stresses in a LEU metal foil annular target.

The effects of fission gas pressure, uranium swelling and thermal contact conductance on the thermal-mechanical behavior of an annular target containing a low-enriched uranium foil (LEU) encapsulated in a nickel foil have been presented in this paper. The draw-plug assembly method is simulated to obtain the residual stresses, which are applied to the irradiation model as initial inputs, and the integrated assembly-irradiation process is simulated as an axisymmetric problem using the commercial finite element code Abaqus FEA. Parametric studies were performed on the LEU heat generation rate and the results indicate satisfactory irradiation performance of the annular target. The temperature and stress margins have been provided along with a discussion of the results.

Future U.S. supply of Mo-99 production through fission based LEU/LEU technology.

Coquí RadioPharmaceuticals Corp. (Coquí) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coquí will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coquí identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coquí by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020.

Production of medical radionuclides in Russia: status and future--a review.

We present a review of reactor and accelerator centers in Russia that produce medical isotopes, the majority of which are exported. In the near future, we anticipate increased isotope production for use in nuclear medicine in Russia. The existing linear accelerator at the Institute for Nuclear Research (Moscow-Troitsk) and several prospective installations are considered to be particularly capable of providing mass production of radionuclides that can substitute, to a certain extent, for the traditional medical isotopes.

Abatement of xenon and iodine emissions from medical isotope production facilities.

The capability of the International Monitoring System (IMS) to detect xenon from underground nuclear explosions is dependent on the radioactive xenon background. Adding to the background, medical isotope production (MIP) by fission releases several important xenon isotopes including xenon-133 and iodine-133 that decays to xenon-133. The amount of xenon released from these facilities may be equivalent to or exceed that released from an underground nuclear explosion. Thus the release of gaseous fission products within days of irradiation makes it difficult to distinguish MIP emissions from a nuclear explosion. In addition, recent shortages in molybdenum-99 have created interest and investment opportunities to design and build new MIP facilities in the United States and throughout the world. Due to the potential increase in the number of MIP facilities, a discussion of abatement technologies provides insight into how the problem of emission control from MIP facilities can be tackled. A review of practices is provided to delineate methods useful for abatement of medical isotopes.

Evaluación de riesgos asociados a la producción de generadores de Molibdeno-99/Tecnecio-99m / Risk analysis applied to the production of generators of Molybdenum-99/Technetium-99m

El Tecnecio-99m es el radionúclido más utilizado en la actualidad en el campo de la medicina nuclear. El Centro Nacional de Seguridad Nuclear (CNSN) como órgano regulador de Cuba requiere la realización de análisis de riesgo a las instalaciones con peligro radiológico potencial asociado y preferentemente, que estos análisis sean realizados mediante el método de matriz de riesgo. Aunque existen varios métodos cualitativos o cuantitativos para realizar estos estudios, la existencia de bases de riesgo de partida y de códigos de análisis, así como la simplicidad de su empleo han privilegiado la matriz de riesgo como método semicuantitativo predominante. El artículo presenta los resultados de la aplicación de este método para la evaluación de los riesgos asociados a la producción de generadores de Molibdeno-99/Tecnecio-99m, que se realiza en el Centro de Isótopos (CENTIS). Empleando el código SECURE-MR Ver. 2.0 se realiza el análisis de riesgo detallado de la práctica. Se identificaron tres secuencias accidentales con nivel de riesgo alto y se determinó la importancia de barreras y reductores relacionados con factores humanos por suficiencia y capacitación de personal. Entre las aplicaciones derivadas del estudio se reportaron capacidades novedosas de análisis, así como posibilidades de monitoreo del riesgo. El modelo de riesgo desarrollado para el proceso de producción de generadores de Molibdeno-99/Tecnecio-99m unido a las capacidades de revisión, análisis y documentación ofrecidas por la herramienta permitieron la obtención de un documento coherente con el modelo, que constituye, además, una base de conocimientos efectiva para la evaluación de la seguridad de la instalación

Technetium-99m is the most commonly used radionuclide in the field of nuclear medicine. The National Center for Nuclear Safety (NCNS), as the regulatory body of Cuba, requires the carrying out of risk analysis for facilities with associated radiological hazard potential and, preferably, that these analyzes be performed using the risk matrix method. Although there are several qualitative or quantitative methods to perform these studies, the existence of basis risk starting and analysis codes and simplicity of use has been privileged the risk matrix as a predominant semi-quantitative method. The paper presents the results of the application of this method for the evaluation of the risks associated with the production of Molybdenum-99/Technetium-99m generators, which is carried out at the Isotope Center (Centis). Using the SECURE - MR Ver. 2.0 code a detailed risk analysis of its practice was carried out. Three accidental sequences of high-level risks were identified and the importance of barriers and reducers related to human factors associated to sufficiency and staff training was determined. Among the derived applications novel analysis capacities as well as risk monitoring possibilities are reported. The risk model developed for the production process of Molybdenum-99/Technetium-99m generators, joined with the ability to review, analyze and all the documentation provided by the tool, allow obtaining a completely and coherent document with the model, which constitutes an effective basis the safety of the facility