Search results for: radio-isotope

Authors: Youngmok Yun, Bosung Kim, Sungho Lee, Kyeongsu Jeon, Hyunwook Hwangbo, Byounggun Choi

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A betavoltaic battery converts nuclear energy released as beta particles (β-) directly into electrical energy. Betavoltaic cells are analogous to photovoltaic cells. The beta particle’s kinetic energy enters a p-n junction and creates electron-hole pairs. Subsequently, the built-in potential of the p-n junction accelerates the electrons and ions to their respective collectors. The major challenges are electrical conversion efficiencies and exact evaluation. In this study, the performance of betavoltaic battery was evaluated. The betavoltaic cell was evaluated in the same condition as radiation from radioactive isotope using by FE-SEM(field emission scanning electron microscope). The average energy of the radiation emitted from the Ni-63 radioisotope is 17.42 keV. FE-SEM is capable of emitting an electron beam of 1-30keV. Therefore, it is possible to evaluate betavoltaic cell without radioactive isotopes. The betavoltaic battery consists of radioisotope that is physically connected on the surface of Si-based PN diode. The performance of betavoltaic battery can be estimated by the efficiency of PN diode unit cell. The current generated by scanning electron microscope with fixed accelerating voltage (17keV) was measured by using faraday cup. Electrical characterization of the p-n junction diode was performed by using Nano Probe Work Station and I-V measurement system. The output value of the betavoltaic cells developed by this research team was 0.162 μw/cm2 and the efficiency was 1.14%.

Keywords: betavoltaic, nuclear, battery, Ni-63, radio-isotope

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Authors: O. Maranci, A. B. Tugrul

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Scintigraphy is an imaging method of nuclear events provoked by collisions or charged current interactions with radiation. It is used for diagnostic test used in nuclear medicine via radiopharmaceuticals emitting radiation which is captured by gamma cameras to form two-dimensional images. Liver scintigraphy is widely used in nuclear medicine.Tc-99m and Cr-51 gamma radioisotopes can be used for this purpose. Cr-51 usage is more important for patients’ organ dose that has higher energy and longer half-life as compared to Tc-99m. In this study, it is aimed to determine the required dose for critical organs of patient through liver scintigraphy via Cr-51 gamma radioisotope. Experimental studies were conducted on patients even though conducting experimental studies on patients is extremely difficult for determination of critical organ doses. Torso phantom was utilized to simulate the liver scintigraphy by using 20 mini packages of Cr-51 that were placed on the organ. The radioisotope was produced by irradiation in central thimble of TRIGA MARK II Reactor at 250 KW power. As the results of the study, critical organ doses were determined and evaluated with different critic organs.

Keywords: critical organ doses, liver, scintigraphy, TRIGA Mark-II

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Authors: Bosung Kim, Youngmok Yun, Sungho Lee, Chanseok Park

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There is a limitation to power supply and operation by the chemical or physical battery in the space environment. Therefore, research for utilizing nuclear energy in the universe has been in progress since the 1950s, around the major industrialized countries. In this study, the self-rechargeable battery having a long life relative to the half-life of the radioisotope is suggested. The hybrid system is composed of betavoltaic battery, all solid battery and energy harvesting board. Betavoltaic battery can produce electrical power at least 10 years over using the radioisotope from Ni-63 and the silicon-based semiconductor. The electrical power generated from the betavoltaic battery is stored in the all-solid battery and stored power is used if necessary. The hybrid system board is composed of input terminals, boost circuit, charging terminals and output terminals. Betavoltaic and all solid batteries are connected to the input and output terminal, respectively. The electric current of 10 µA is applied to the system board by using the high-resolution power simulator. The system efficiencies are measured from a boost up voltage of 1.8 V, 2.4 V and 3 V, respectively. As a result, the efficiency of system board is about 75% after boosting up the voltage from 1V to 3V.

Keywords: isotope, betavoltaic, nuclear, battery, energy harvesting

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Authors: Alex B. Cusick

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The objective of this study is to improve upon the current ²³⁸PuO₂ fuel technology for space and defense applications. Modern RTGs (radioisotope thermoelectric generators) utilize the heat generated from the radioactive decay of ²³⁸Pu to create heat and electricity for long term and remote missions. Application of RTG technology is limited by the scarcity and expense of producing the isotope, as well as the power output which is limited to only a few hundred watts. The scarcity and expense make the efficient use of ²³⁸Pu absolutely necessary. By utilizing the decay of ²³⁸Pu, not only to produce heat directly but to also indirectly induce fission in ²³⁹Pu (which is already present within currently used fuel), it is possible to see large increases in temperature which allows for a more efficient conversion to electricity and a higher power-to-weight ratio. This concept can reduce the quantity of ²³⁸Pu necessary for these missions, potentially saving millions on investment, while yielding higher power output. Current work investigating radioisotope power systems have focused on improving efficiency of the thermoelectric components and replacing systems which produce heat by virtue of natural decay with fission reactors. The technical feasibility of utilizing (α,n) reactions to induce fission within current radioisotopic fuels has not been investigated in any appreciable detail, and our study aims to thoroughly investigate the performance of many such designs, develop those with highest capabilities, and facilitate experimental testing of these designs. In order to determine the specific design parameters that maximize power output and the efficient use of ²³⁸Pu for future RTG units, MCNP6 simulations have been used to characterize the effects of modifying fuel composition, geometry, and porosity, as well as introducing neutron moderating, reflecting, and shielding materials to the system. Although this project is currently in the preliminary stages, the final deliverables will include sophisticated designs and simulation models that define all characteristics of multiple novel RTG fuels, detailed enough to allow immediate fabrication and testing. Preliminary work has consisted of developing a benchmark model to accurately represent the ²³⁸PuO₂ pellets currently in use by NASA; this model utilizes the alpha transport capabilities of MCNP6 and agrees well with experimental data. In addition, several models have been developed by varying specific parameters to investigate their effect on (α,n) and (n,fi ssion) reaction rates. Current practices in fuel processing are to exchange out the small portion of naturally occurring ¹⁸O and ¹⁷O to limit (α,n) reactions and avoid unnecessary neutron production. However, we have shown that enriching the oxide in ¹⁸O introduces a sufficient (α,n) reaction rate to support significant fission rates. For example, subcritical fission rates above 10⁸ f/cm³-s are easily achievable in cylindrical ²³⁸PuO₂ fuel pellets with a ¹⁸O enrichment of 100%, given an increase in size and a ⁹Be clad. Many viable designs exist and our intent is to discuss current results and future endeavors on this project.

Keywords: radioisotope thermoelectric generators (RTG), Pu-238, subcritical reactors, (alpha, n) reactions

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Authors: C. G. Mendez-Garcia, E. T. Romero-Guzman, H. Hernandez-Mendoza, C. Solis, E. Chavez-Lomeli, E. Chamizo, R. Garcia-Tenorio

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Plutonium is present in different concentrations in the environment and biological samples related to nuclear weapons testing, nuclear waste recycling and accidental discharges of nuclear plants. This radioisotope is considered the most radiotoxic substance, particularly when it enters the human body through inhalation of powders insoluble or aerosols. This is the main reason of the determination of the concentration of this radioisotope in the atmosphere. Besides that, the isotopic ratio of ²⁴⁰Pu/²³⁹Pu provides information about the origin of the source. PM₂.₅ sampling was carried out in the Metropolitan Zone of the Valley of Mexico (MZVM) from February 18th to March 17th in 2015 on quartz filter. There have been significant developments recently due to the establishment of new methods for sample preparation and accurate measurement to detect ultra trace levels as the plutonium is found in the environment. The accelerator mass spectrometry (AMS) is a technique that allows measuring levels of detection around of femtograms (10-15 g). The AMS determinations include the chemical isolation of Pu. The Pu separation involved an acidic digestion and a radiochemical purification using an anion exchange resin. Finally, the source is prepared, when Pu is pressed in the corresponding cathodes. According to the author's knowledge on these aerosols showed variations on the ²³⁵U/²³⁸U ratio of the natural value, suggesting that could be an anthropogenic source altering it. The determination of the concentration of the isotopes of Pu can be a useful tool in order the clarify this presence in the atmosphere. The first results showed a mean value of activity concentration of ²³⁹Pu of 280 nBq m⁻³ thus the ²⁴⁰Pu/²³⁹Pu was 0.025 corresponding to the weapon production source; these results corroborate that there is an anthropogenic influence that is increasing the concentration of radioactive material in PM₂.₅. According to the author's knowledge in Total Suspended Particles (TSP) have been reported activity concentrations of ²³⁹⁺²⁴⁰Pu around few tens of nBq m⁻³ and 0.17 of ²⁴⁰Pu/²³⁹Pu ratios. The preliminary results in MZVM show high activity concentrations of isotopes of Pu (40 and 700 nBq m⁻³) and low ²⁴⁰Pu/²³⁹Pu ratio than reported. These results are in the order of the activity concentrations of Pu in weapons-grade of high purity.

Keywords: aerosols, fallout, mass spectrometry, radiochemistry, tracer, ²⁴⁰Pu/²³⁹Pu ratio

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Authors: Evgeny E. Tereshatov

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Astatine-211 is considered one of the most promising radionuclides for Targeted Alpha Therapy. In order to develop reliable procedures to label biomolecules and utilize efficient delivery vehicle principles, one should understand the main chemical characteristics of astatine. The short half-life of 211At (~7.2 h) and absence of any stable isotopes of this element are limiting factors towards studying the behavior of astatine. Our team has developed a procedure for rapid and efficient isolation of astatine from irradiated bismuth material in nitric acid media based on 3-octanone and 1-octanol extraction chromatography resins. This process has been automated and it takes 20 min from the beginning of the target dissolution to the At-211 fraction elution. Our next step is to consider commercially available chromatography resins and their applicability in astatine purification in the same media. Results obtained along with the corresponding sorption mechanisms will be discussed.

Keywords: astatine-211, chromatography, automation, mechanism, radiopharmaceuticals

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Authors: Gulten Yilmaz, A. Beril Tugrul, Mustafa Demir, Dogan Yasar, Bayram Demir, Bulent Buyuk

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A Positron Emission Tomography (PET) is a radioisotope imaging technique that illustrates the organs and the metabolisms of the human body. This technique is based on the simultaneous detection of 511 keV annihilation photons, annihilated as a result of electrons annihilating positrons that radiate from positron-emitting radioisotopes that enter biological active molecules in the body. This study was conducted on ten patients in an effort to conduct patient-related experimental studies. Dosage monitoring for the bladder, which was the organ that received the highest dose during PET applications, was conducted for 24 hours. Assessment based on measuring urination activities after injecting patients was also a part of this study. The MIRD method was used to conduct dosage calculations for results obtained from experimental studies. Results obtained experimentally and theoretically were assessed comparatively.

Keywords: PET/CT, TLD, MIRD, dose measurement, patient doses

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Authors: Hadi Shamoradifar, Behzad Teimuri, Parviz Parvaresh, Saeed Mohammadi

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One of the experimental facilities of the heavy water research reactor is the central test loop (C.T.L). It is located along the central axial line of the vessel, and therefore will highly affect the neutronic parameters of the reactor, so from the neutronics point of view, C.T.L is the most important facility. It is mainly designed for fuel testing, thought other applications such as radioisotope production and neutron activation, can be imagine for it. All of the simulations were performed by MCNPX2.6. As a first step towards C.T.L analysis, the effect of D2O-filled, H2O-filled, and He-filled C.T.L on the effective multiplication factor (Keff.), have been evaluated. According to results, H2O-filled C.T.L has a higher thermal neutron, while He-filled C.T.L includes more resonance neutrons. In the next step thermal and total axial neutron fluxes, were calculated and used as the comparison parameters. The core without C.T.L (C.T.L replaced by heavy water) is selected as the reference case, and the effect of all other cases is calculated according to that.

Keywords: heavy water reactor, neutronic calculations, central test loop, neutron activation

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Authors: Rian O. Miranda, Lidia V. de Sa, Julio C. Suita

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The Brazilian Nuclear Energy Commission (CNEN) is the main manufacturer of radiopharmaceuticals in Brazil. The Nuclear Engineering Institute (IEN), located at Rio de Janeiro, is one of its main centers of research and production, attending public and private hospitals in the state. This radiopharmaceutical production is used in diagnostic and therapy procedures and allows one and a half million nuclear medicine procedures annually. Despite this, the country is not self-sufficient to meet national demand, creating the need for importation and consequent dependence on other countries. However, IEN facilities were designed in the 60's, and today its structure is inadequate in relation to the good manufacturing practices established by sanitary regulator (ANVISA) and radiological protection leading to the need for a new project. In order to adapt and increase production in the country, a new plant will be built and integrated to the existing facilities with a new 30 MeV Cyclotron that is actually in project detailing process. Thus, it is proposed to survey current CNEN and ANVISA standards for radiopharmaceutical production facilities, as well as the radiological protection analysis of each area of the plant, following good manufacturing practices recommendations adopted nationally besides licensing exigencies for radioactive facilities. In this way, the main requirements for proper operation, equipment location, building materials, area classification, and maintenance program have been implemented. The access controls, interlocks, segregation zones and pass-through boxes integrated into the project were also analyzed. As a result, IEN will in future have the flexibility to produce all necessary radioisotopes for nuclear medicine application, more efficiently by simultaneously bombarding two targets, allowing the simultaneous production of two different radioisotopes, minimizing radiation exposure and saving operating costs.

Keywords: cyclotron, legislation, norms, production, radiopharmaceuticals

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Authors: H. Duggal, G. Singh, G. Singh, A. Bhalla, S. Kumar, J. S. Shahi, D. Mehta

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The total differential cross section for scattering of the 5.895 keV photons by various polymers has been measured at scattering angle of 135o. The experimental measurements were carried out using the energy dispersive setup involving annular source of the 55Fe radioisotope and a low energy germanium (LEGe) detector. The cross section values are measured for 20 polymer targets namely, Paraffin Wax, Polytetrafluoro ethylene (PTFE), Cellulose, Silicone oil, Polyvinyl alcohol (PVA), Polyvinyl purrolidone (PVP), Polymethyl methacrylate (PMMA), Kapton, Mylar, Chitosan, Polyvinyl chloride (PVC), Bakelite, Carbopol, Chlorobutyl rubber (CBR), Polyetylene glycol (PEG), Polysorbate-20, Nylon-6, Cetyl alcohol, Carboxyl methyl sodium cellulose and Sodium starch glucolate. The measurements were performed in vacuum so as to avoid scattering contribution due to air and strong absorption of low energy photons in the air column. In the present investigations, the geometrical factor and efficiency of the detector were determined by measuring the K x-rays emitted from the 22Ti and 23V targets excited by the Mn K x-rays in the same experimental set up. The measured scattering cross sections have been compared with the sum of theoretically calculated elastic and inelastic scattering cross sections. The theoretical elastic (Rayleigh) scattering cross sections based on the various form factor approximations, namely, non-relativistic form factor (NF), relativistic form factor (RF), modified form factor (MF), and MF with anomalous scattering factor (ASF) as well as the second order S-matrix formalisms, and the inelastic scattering differential cross sections based on the Klein-Nishina formula after including the inelastic scattering function (KN+ISF) have been calculated. The experimental results show fairly good agreement with theoretical cross sections.

Keywords: photon, polymers, elastic and inelastic, scattering cross sections

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Authors: Yeo Ryeong Jeon, Pyong Kon Cho, Eun Ok Han, Hyon Chul Jang, Yong Min Kim

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This study surveyed on recognition related to radiation safety for radiation safety managers and workers those who have been worked in Republic of Korea education and research institution. At present, South Korea has no guideline and manual of radiation safety for education and research institution. Therefore, we tried to find an educational basis for development of radiation safety guideline and manual. To check the level of knowledge, attitude, and behavior about radiation safety, we used the questionnaire that consisted of 29 questions against knowledge, attitude and behavior, 4 questions against self-efficacy and expectation based on four factors (radiation source, human, organizational and physical environment) of the Haddon's matrix. Responses were collected between May 4 and June 30, 2015. We analyzed questionnaire by means of IBM SPSS/WIN 15 which well known as statistical package for social science. The data were compared with mean, standard deviation, Pearson's correlation, ANOVA (analysis of variance) and regression analysis. 180 copies of the questionnaire were returned from 60 workplaces. The overall mean results for behavior level was relatively lower than knowledge and attitude level. In particular, organizational environment factor on the radiation safety management indicated the lowest behavior level. Most of the factors were correlated in Pearson’s correlation analysis, especially between knowledge of human factors and behavior of human factors (Pearson’s correlation coefficient 0.809, P<.01). When analysis performed in line with the main radiation source type, institutions where have been used only opened RI (radioisotope) behavior level was the lowest among all subjects. Finally, knowledge of radiation source factor (β=0.556, P<.001) and human factor(β=0.376, P<.001) had the greatest impact in terms of behavior practice. Radiation safety managers and workers think positively about radiation safety management, but are poorly informed organizational environment of their institution. Thus, each institution need to efforts to settlement of radiation safety culture. Also, pedagogical interventions for improving knowledge on radiation safety needs in terms of safety accident prevention.

Keywords: radiation safety management, factor analysis, SPSS, republic of Korea

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Authors: Radu Deju, Carmen Mustata

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The VVR-S nuclear research reactor owned by Horia Hubulei National Institute of Physics and Nuclear Engineering (IFIN-HH) was designed for research and radioisotope production, being permanently shut-down in 2002, after 40 years of operation. All amount of the nuclear spent fuel S-36 and EK-10 type was returned to Russian Federation (first in 2009 and last in 2012), and the radioactive waste resulted from the reprocessing of it will remain permanently in the Russian Federation. The decommissioning strategy chosen is immediate dismantling. At this moment, the radionuclides with half-life shorter than 1 year have a minor contribution to the contamination of materials and equipment used in reactor department. The decommissioning of the reactor has started in 2010 and is planned to be finalized in 2020, being the first nuclear research reactor that has started the decommissioning project from the South-East of Europe. The management system applied in the decommissioning of the VVR-S research reactor integrates all common elements of management: nuclear safety, occupational health and safety, environment, quality- compliance with the requirements for decommissioning activities, physical protection and economic elements. This paper presents the application of integrated management system in decommissioning of systems, structures, equipment and components (SSEC) from pumps room, including the management of the resulted radioactive waste. The primary cooling system of this type of reactor includes circulation pumps, heat exchangers, degasser, filter ion exchangers, piping connection, drainage system and radioactive leaks. All the decommissioning activities of primary circuit were performed in stage 2 (year 2014), and they were developed and recorded according to the applicable documents, within the requirements of the Regulatory Body Licenses. In the presentation there will be emphasized how the integrated management system provisions are applied in the dismantling of the primary cooling system, for elaboration, approval, application of necessary documentation, records keeping before, during and after the dismantling activities. Radiation protection and economics are the key factors for the selection of the proper technology. Dedicated and advanced technologies were chosen to perform specific tasks. Safety aspects have been taken into consideration. Resource constraints have also been an important issue considered in defining the decommissioning strategy. Important aspects like radiological monitoring of the personnel and areas, decontamination, waste management and final characterization of the released site are demonstrated and documented.

Keywords: decommissioning, integrated management system, nuclear reactor, waste management

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