Search results for: radioactive fission products

Authors: Kobi Kutsher

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On February 12th 2013, monitoring stations of the Preparatory Commission of the Comprehensive Nuclear Test-Ban Treaty Organization (CTBTO) detected a seismic event with explosion-like underground characteristics in the Democratic People’s Republic of Korea (DPRK). The location was found to be in the vicinity of the two previous announced nuclear tests in 2006 and 2009.The nuclear test was also announced by the government of the DPRK.After an underground nuclear explosion, radioactive fission products (mostly noble gases) can seep through layers of rock and sediment until they escape into the atmosphere. The fission products are dispersed in the atmosphere and may be detected thousands of kilometers downwind from the test site. Indeed, more than 7 weeks after the explosion, unusual detections of noble gases was reported at the radionuclide station in Takasaki, Japan. The radionuclide station is a part of the International Monitoring System, operated to verify the CTBT. This study provides an estimation of the possible source region and the total radioactivity of the release using Atmospheric Transport Modeling.

Keywords: atmospheric transport modeling, CTBTO, nuclear tests, radioactive fission products

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Authors: Haiming Wen, Isabella J. Van Rooyen

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Tristructural isotropic (TRISO)-coated fuel particles are designed as nuclear fuel for high-temperature gas reactors. TRISO coating consists of layers of carbon buffer, inner pyrolytic carbon (IPyC), SiC, and outer pyrolytic carbon. The TRISO coating, especially the SiC layer, acts as a containment system for fission products produced in the kernel. However, release of certain metallic fission products across intact TRISO coatings has been observed for decades. Despite numerous studies, mechanisms by which fission products migrate across the coating layers remain poorly understood. In this study, scanning transmission electron microscopy (STEM), energy dispersive X-ray spectroscopy (EDS), high-resolution transmission electron microscopy (HRTEM) and electron energy loss spectroscopy (EELS) were used to examine the distribution, composition and structure of fission products in a TRISO coated particle neutron irradiated to 3.6 x 1021 n/cm² fast fluence at 1040 ⁰C. Precession electron diffraction was used to investigate characters of grain boundaries where specific fission product precipitates are located. The retention fraction of 110mAg in the investigated TRISO particle was estimated to be 0.19. A high density of nanoscale fission product precipitates was observed in the SiC layer close to the SiC-IPyC interface, most of which are rich in Pd, while Ag was not identified. Some Pd-rich precipitates contain U. Precipitates tend to have complex structure and composition. Although a precipitate appears to have uniform contrast in STEM, EDS indicated that there may be composition variations throughout the precipitate, and HRTEM suggested that the precipitate may have several parts different in crystal structure or orientation. Attempts were made to measure charge states of precipitates using EELS and study their possible effect on precipitate transport.

Keywords: TRISO particle, fission product, nuclear fuel, electron microscopy, neutron irradiation

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Authors: Hayrettin Eroglu, Oguz Aksakal

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In this study, it is analysed that which decontamination products are more effective and how decontamination process should be performed in the case of contamination of radioactive 99mTc which is the most common radioactive element used in nuclear applications dealing with the human body or the environment. Based on the study, it is observed that existing radioactive washers are less effective than expected, alcohol has no effect on the decontamination of 99mTc, and temperature and pH are the most important factors. In the light of the analysis, it is concluded that the most effective decontamination product is DM-D (Decontamination Material-D). When the effect of DM-D on surfaces is analysed, it is observed that decontamination is very fast on scrubs and formica with both DM-D and water, and although DM-D is very effective on skin, it is not effective on f ceramic tiles and plastic floor covering material. Also in this study, the effectiveness of different molecular groups in the decontaminant was investigated. As a result, the acetate group has been observed as the most effective component of the decontaminant.

Keywords: contamination, radioactive, technetium, decontamination

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Authors: Tuğçe Yavaş Akış

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In recent years, due to an increased use of radioactive materials, radioactive sources are constantly being transported via air, road and ocean ways for medical, industrial, research etc. purposes throughout the world. The quantity of radioactive materials transported all around the world varies from negligible quantities in shipments of consumer products to very large quantities in shipments of irradiated nuclear fuel. Radioactive materials have been less attractive for social science researchers in literature. In this study, it is aimed to discover desired flow of radioactive materials from logistics service quality (LSQ) perspective. In doing so, case study approach will be employed by using secondary data collected from one of the world’s leading transportation companies’ customer care system reports. Movement of radioactive cargoes containing IR-192 and logistics process will be analyzed with the help of logistics service quality dimensions. Based on the case study that will be conducted, interaction between dimensions, the importance of each dimension in desired flow, and their relevance with desired flow of radioactive materials will be explained. This study will bring out the desired flow of radioactive materials transportation and be a guide for all other companies, employees and researchers.

Keywords: logistics service quality, LSQ dimension , radioactive material, transportation

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Authors: Ankita Jadon, Sidi Souvi, Nathalie Girault, Denis Petitprez

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Safety requirements for Generation IV nuclear reactor designs, especially the new generation sodium-cooled fast reactors (SFR) require a risk-informed approach to model severe accidents (SA) and their consequences in case of outside release. In SFRs, aerosols are produced during a core disruptive accident when primary system sodium is ejected into the containment and burn in contact with the air; producing sodium aerosols. One of the key aspects of safety evaluation is the in-containment sodium aerosol behavior and their interaction with fission products. The study of the effects of sodium fires is essential for safety evaluation as the fire can both thermally damage the containment vessel and cause an overpressurization risk. Besides, during the fire, airborne fission product first dissolved in the primary sodium can be aerosolized or, as it can be the case for fission products, released under the gaseous form. The objective of this work is to study the interactions between sodium aerosols and fission products (Iodine, toxic and volatile, being the primary concern). Sodium fires resulting from an SA would produce aerosols consisting of sodium peroxides, hydroxides, carbonates, and bicarbonates. In addition to being toxic (in oxide form), this aerosol will then become radioactive. If such aerosols are leaked into the environment, they can pose a danger to the ecosystem. Depending on the chemical affinity of these chemical forms with fission products, the radiological consequences of an SA leading to containment leak tightness loss will also be affected. This work is split into two phases. Firstly, a method to theoretically understand the kinetics and thermodynamics of the heterogeneous reaction between sodium aerosols and fission products: I2 and HI are proposed. Ab-initio, density functional theory (DFT) calculations using Vienna ab-initio simulation package are carried out to develop an understanding of the surfaces of sodium carbonate (Na2CO3) aerosols and hence provide insight on its affinity towards iodine species. A comprehensive study of I2 and HI adsorption, as well as bicarbonate formation on the calculated lowest energy surface of Na2CO3, was performed which provided adsorption energies and description of the optimized configuration of adsorbate on the stable surface. Secondly, the heterogeneous reaction between (I2)g and Na2CO3 aerosols were investigated experimentally. To study this, (I2)g was generated by heating a permeation tube containing solid I2, and, passing it through a reaction chamber containing Na2CO3 aerosol deposit. The concentration of iodine was then measured at the exit of the reaction chamber. Preliminary observations indicate that there is an effective uptake of (I2)g on Na2CO3 surface, as suggested by our theoretical chemistry calculations. This work is the first step in addressing the gaps in knowledge of in-containment and atmospheric source term which are essential aspects of safety evaluation of SFR SA. In particular, this study is aimed to determine and characterize the radiological and chemical source term. These results will then provide useful insights for the developments of new models to be implemented in integrated computer simulation tool to analyze and evaluate SFR safety designs.

Keywords: iodine adsorption, sodium aerosols, sodium cooled reactor, DFT calculations, sodium carbonate

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Authors: Liu De-jun, Fu Jing, Zhang Rong, Luo Tian, Ma Ning

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Cs-137, the radioactive fission products of uranium, can be easily dissolved in water during the accident of nuclear power plant, such as Chernobyl, Three Mile Island, Fukushima accidents. The concentration of Cs in the groundwater around the nuclear power plant exceeded the standard value almost 10,000 times after the Fukushima accident. The adsorption capacity of Titanate nano-materials for radioactive cation (Cs+) is very strong. Moreover, the radioactive ion can be tightly contained in the nanotubes or nanofibers without reversible adsorption, and it can safely be fixed. In addition, the nano-material has good chemical stability, thermal stability and mechanical stability to minimize the environmental impact of nuclear waste and waste volume. The preparation of titanate nanotubes or nanofibers was studied by hydrothermal methods, and chemical kinetics of removal of Cs by nano-materials was obtained. The adsorption time with maximum adsorption capacity and the effects of pH, coexisting ion concentration and the optimum adsorption conditions on the removal of Cs by titanate nano-materials were also obtained. The adsorption boundary curves, adsorption isotherm and the maximum adsorption capacity of Cs-137 as tracer on the nano-materials were studied in the research. The experimental results showed that the removal rate of Cs-137 in 0.01 tons of waste water with only 1 gram nano-materials could reach above 98%, according to the optimum adsorption conditions.

Keywords: preparation, titanate, cs-137, removal, nuclear

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Authors: Masaumi Nakahara, Sou Watanabe, Hiromichi Ogi, Atsuhiro Shibata, Kazunori Nomura

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In the Japan Atomic Energy Agency, three types of experimental research, advanced reactor fuel reprocessing, radioactive waste disposal, and nuclear fuel cycle technology, have been carried out at the Chemical Processing Facility. The facility has generated high level radioactive liquid and solid wastes in hot cells. The high level radioactive solid waste is divided into three main categories, a flammable waste, a non-flammable waste, and a solid reagent waste. A plastic product is categorized into the flammable waste and molten with a heating mantle. The non-flammable waste is cut with a band saw machine for reducing the volume. Among the solid reagent waste, a used adsorbent after the experiments is heated, and an extractant is decomposed for its stabilization. All high level radioactive solid wastes in the hot cells are packed in a high level radioactive solid waste can. The high level radioactive solid waste can is transported to the 2nd High Active Solid Waste Storage in the Tokai Reprocessing Plant in the Japan Atomic Energy Agency.

Keywords: high level radioactive solid waste, advanced reactor fuel reprocessing, radioactive waste disposal, nuclear fuel cycle technology

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Authors: Zhibin Lei, Minghui Yang

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Ningwu-Jingle basin is a structural syncline which has undergone a complex tectono-thermal history since Cretaceous. It stretches along the strike of the northern Lvliang Mountains which are the most important mountains in the middle and west of North China Craton. The Mesozoic units make up of the core of Ningwu-Jingle Basin, with pre-Mesozoic units making up of its flanks. The available low-temperature thermochronology implies that Ningwu-Jingle Basin has experienced two stages of uplifting: 94±7Ma to 111±8Ma (Albian to Cenomanian) and 62±4 to 75±5Ma (Danian to Maastrichtian). In order to constrain its tectono-thermal history in the Cenozoic, both apatite (U-Th-Sm)/He and fission track dating analysis are applied on 3 Middle Jurassic and 3 Upper Triassic sandstone samples. The central fission track ages range from 74.4±8.8Ma to 66.0±8.0Ma (Campanian to Maastrichtian) which matches well with previous data. The central He ages range from 20.1±1.2Ma to 49.1±3.0Ma (Ypresian to Burdigalian). Inverse thermal modeling is established based on both apatite fission track data and (U-Th-Sm)/He data. The thermal history obtained reveals that all 6 sandstone samples cross the high-temperature limit of fission track partial annealing zone by the uppermost Cretaceous and that of He partial retention zone by the uppermost Eocene to the early Oligocene. The result indicates that the middle and west of North China Craton is not stable in the Cenozoic.

Keywords: apatite fission track thermochronology, apatite (u–th)/he thermochronology, Ningwu-Jingle basin, North China craton, tectono-thermal history

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Authors: Ilija Plecas, Dalibor Arbutina

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The leaching rate of 60Co from spent mix bead (anion and cation) exchange resins in a cement-bentonite matrix has been studied. Transport phenomena involved in the leaching of a radioactive material from a cement-bentonite matrix are investigated using three methods based on theoretical equations. These are: the diffusion equation for a plane source, an equation for diffusion coupled to a first order equation and an empirical method employing a polynomial equation. The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choices for radioactive waste packaging for a future Serbian radioactive waste disposal center.

Keywords: cement, concrete, immobilization, leaching, permeability, radioactivity, waste

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Authors: Sara Mostofian

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The nuclear industry is a technology that can fulfill future energy needs but requires special attention to ensure safety and reliability while minimizing any environmental impact. To meet these expectations, the nuclear industry is exploring different reactor technologies for power production. Several designs are under development and the technical viability of these new designs is the subject of many ongoing studies. One of these studies considers the radioactive emissions and radioactive waste generated during the life of a nuclear power production plant to allow a successful license process. For all the modern technologies, a good understanding of the radioactivity generated in the process systems of the plant is essential. Some of that understanding may be gleaned from the performance of some prototype reactors of similar design that operated decades ago. This paper presents how, with that understanding, a model can be developed to estimate the emissions as well as the radioactive waste during the normal operation of a nuclear power plant. The model would predict the radioactive material concentrations in different waste streams. Using this information, the radioactive emission and waste generated during the life of these new technologies can be estimated during the early stages of the design of the plant.

Keywords: SMRs, activity transport, model, radioactive waste

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Authors: S. Sandri, G. M. Contessa, C. Poggi

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An experimental nuclear fusion device could be seen as a step toward the development of the future nuclear fusion power plant. If compared with other possible solutions to the energy problem, nuclear fusion has advantages that ensure sustainability and security. In particular considering the radioactivity and the radioactive waste produced, in a nuclear fusion plant the component materials could be selected in order to limit the decay period, making it possible the recycling in a new reactor after about 100 years from the beginning of the decommissioning. To achieve this and other pertinent goals many experimental machines have been developed and operated worldwide in the last decades, underlining that radiation protection and workers exposure are critical aspects of these facilities due to the high flux, high energy neutrons produced in the fusion reactions. Direct radiation, material activation, tritium diffusion and other related issues pose a real challenge to the demonstration that these devices are safer than the nuclear fission facilities. In Italy, a limited number of fusion facilities have been constructed and operated since 30 years ago, mainly at the ENEA Frascati Center, and the radiation protection approach, addressed by the national licensing requirements, shows that it is not always easy to respect the constraints for the workers' exposure to ionizing radiation. In the current analysis, the main radiation protection issues encountered in the Italian Fusion facilities are considered and discussed, and the technical and legal requirements are described. The licensing process for these kinds of devices is outlined and compared with that of other European countries. The following aspects are considered throughout the current study: i) description of the installation, plant and systems, ii) suitability of the area, buildings, and structures, iii) radioprotection structures and organization, iv) exposure of personnel, v) accident analysis and relevant radiological consequences, vi) radioactive wastes assessment and management. In conclusion, the analysis points out the needing of a special attention to the radiological exposure of the workers in order to demonstrate at least the same level of safety as that reached at the nuclear fission facilities.

Keywords: fusion facilities, high energy neutrons, licensing process, radiation protection

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Authors: V. Thomas, O. Delaune, W. Hennig, S. Hoover

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Measurement of radioactive isotopes of atmospheric xenon is used to detect, locate and identify any confined nuclear tests as part of the Comprehensive Nuclear Test-Ban Treaty (CTBT). In this context, the Alternative Energies and French Atomic Energy Commission (CEA) has developed a fixed device to continuously measure the concentration of these fission products, the SPALAX process. During its atmospheric transport, the radioactive xenon will undergo a significant dilution between the source point and the measurement station. Regarding the distance between fixed stations located all over the globe, the typical volume activities measured are near 1 mBq m⁻³. To avoid the constraints induced by atmospheric dilution, the development of a mobile detection system is in progress; this system will allow on-site measurements in order to confirm or infringe a suspicious measurement detected by a fixed station. Furthermore, this system will use beta/gamma coincidence measurement technique in order to drastically reduce environmental background (which masks such activities). The detector prototype consists of a gas cell surrounded by two large silicon wafers, coupled with two square NaI(Tl) detectors. The gas cell has a sample volume of 30 cm³ and the silicon wafers are 500 µm thick with an active surface area of 3600 mm². In order to minimize leakage current, each wafer has been segmented into four independent silicon pixels. This cell is sandwiched between two low background NaI(Tl) detectors (70x70x40 mm³ crystal). The expected Minimal Detectable Concentration (MDC) for each radio-xenon is in the order of 1-10 mBq m⁻³. Three 4-channels digital acquisition modules (Pixie-NET) are used to process all the signals. Time synchronization is ensured by a dedicated PTP-network, using the IEEE 1588 Precision Time Protocol. We would like to present this system from its simulation to the laboratory tests.

Keywords: beta/gamma coincidence technique, low level measurement, radioxenon, silicon pixels

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Authors: Hai Huu Le, Paul Junor, Moshi Geso, Graeme O’Keefe

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In this paper, the authors propose a personal wearable directionally-sensitive radiation dosimeter using multiple semiconductor CdZnTe detectors. The proposed dosimeter not only measures the real-time dose rate but also provide the direction of the radioactive source. A linear relationship between radioactive source direction and the radiation intensity measured by each detectors is established and an equation to determine the source direction is derived by the authors. The efficiency and accuracy of the proposed dosimeter is verified by simulation using Geant4 package. Results have indicated that in a measurement duration of about 7 seconds, the proposed dosimeter was able to estimate the direction of a 10μCi 137/55Cs radioactive source to within 2 degrees.

Keywords: dose rate, Geant4 package, radiation dosimeter, radioactive source direction

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Authors: Marisé García Batlle, Juan Manuel Navarrete Tejero

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In aquatic ecosystems, the main indicators of pollution are contaminated sediments, which are the primary repository of radionuclides and chemicals elements in the marine environment. Radioactive Contamination Factor (RCF) has been proposed as a suitable unit to measure the magnitude of radioactive contamination at global scale, caused mainly by more than 2,000 nuclear explosions tests performed during the 1945-65 period. It is obtained as percentage of contaminant radioactivity (¹³⁷Cs) compared to natural radioactivity (⁴⁰K), both expressed in Bq/g of marine sediments conditioned in Marinelli containers and detected in both NaI(Tl) and HPGe detectors. So, in this paper samples of marine sediments were taken up along the occidental Cuban coasts and analyzed by gamma spectrometry for the determination of gamma-emitting radioisotopes with energies between 60 and 2000 keV. The results proved that the proposed method is simple and suitable to evaluated radioactive contamination. Also, the RCF values provide an appropriate indicator to predict which pollution levels in the future will be and if the rate will go down as disintegrates the ¹³⁷Cs present when only 2,4 half-lives have passed away.

Keywords: Cuba, gamma spectrometry, marine sediments, radioactive pollution

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Authors: Myung Ho Lee, Hee Seung Lim, Young Jae Maeng, Chang Hoon Lee

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This study presents an automatic sequential extraction device for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO₃, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO₃ medium with anion exchange resin. For leaching solution passed through the anion exchange column, the Am isotopes were sequentially separated with TRU resin. Automatic sequential extraction device built-in software information of separation for Pu and Am isotopes was developed. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after the micro-precipitation of neodymium. The data of Pu and Am isotopes in radioactive waste with an automatic sequential extraction device developed in this study were validated with the ICP-MS system.

Keywords: automatic sequential extraction device, Pu isotopes, Am isotopes, alpha spectrometer, radioactive waste samples, ICP-MS system

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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: R. Gharari, N. Kojouri, R. Hosseini Aghdam, E. Alibeigi, B. Salmasian

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In this research, the WWER1000/V446 (a PWR Russian type reactor) is chosen as the case study. It is assumed that radioactive materials that release into the environment are more than allowable limit due to a complete failure of the ventilation system (reactor stack). In the following, the HOTSPOT and the RASCAL computational codes have been used and coupled with a developed program using MATLAB software to evaluate Total effective dose equivalent (TEDE) and cancer risk according to the BEIR equations for various human organs. In addition, effects of the containment spray system and climate conditions on the TEDE have been investigated. According to the obtained results, there is an inverse correlation between the received dose and the wind speed; the amount of the TEDE for wind speed 2 m/s and is more than wind speed for 14 m/s during the class A of the climate (2.168 and 0.444 mSv, respectively). Also, containment spray system can effect and reduce the amount of the fission products and TEDE. Furthermore, the probability of the cancer risk for women is more than men, and for children is more than adults. In addition, a specific emergency zonal planning is proposed. Results are promising in which the site selection of the WWER1000/V446 were considered safe for the public in this situation.

Keywords: TEDE, total effective dose equivalent, RASCAL and HOTSPOT codes, BEIR equations, cancer risk

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Authors: Tsai-chun Lai, Szu-ju Fu, Tzu-lin Lee, Yuh-Lien Chen

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There is much evidence that exposure to fine particulate matter (PM) from air pollution increases the risk of cardiovascular morbidity and mortality. According to previous reports, PM in the air enters the respiratory tract, contacts the alveoli, and enters the blood circulation, leading to the progression of cardiovascular disease. PM pollution may also lead to cardiometabolic disturbances, increasing the risk of cardiovascular disease. The effects of PM on cardiac function and mitochondrial damage are currently unknown. We used mice and rat cardiomyocytes (H9c2) as animal and in vitro cell models, respectively, to simulate an air pollution environment using PM. These results indicate that the apoptosis-related factor PUMA, a regulator of apoptosis upregulated by p53, is increased in mice treated with PM. Apoptosis was aggravated in cardiomyocytes treated with PM, as measured by TUNEL assay and Annexin V/PI. Western blot results showed that CASPASE3 was significantly increased and BCL2 (B-cell lymphoid 2) was significantly decreased under PM treatment. Concurrent exposure to PM increases mitochondrial reactive oxygen species (ROS) production by MitoSOX Red staining. Furthermore, using Mitotracker staining, PM treatment significantly shortened mitochondrial length, indicating mitochondrial fission. The expression of mitochondrial fission-related proteins p-DRP1 (phosphodynamics-related protein 1) and FIS1 (mitochondrial fission 1 protein) was significantly increased. Based on these results, the exposure to PM worsens mitochondrial function and leads to cardiomyocyte apoptosis.

Keywords: particulate matter, cardiomyocyte, apoptosis, mitochondria

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Authors: Michael Dole, Pierre Ninin, Denis Raffourt

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Most of the CERN’s facilities hosting particle accelerators are large, underground and radioactive areas. All fire detection systems installed in such areas, shall be carefully studied to cope with the particularities of this stringent environment. The detection equipment usually chosen by CERN to secure these underground facilities are based on air sampling technology. The electronic equipment is located in non-radioactive areas whereas air sampling networks are deployed in radioactive areas where fire detection is required. The air sampling technology provides very good detection performances and prevent the "radiation-to-electronic" effects. In addition, it reduces the exposure to radiations of maintenance workers and is permanently available during accelerator operation. In order to protect the Super Proton Synchrotron and its 7 km tunnels, a specific long distance aspirating smoke detector has been developed to detect smoke at up to 700 meters between electronic equipment and the last air sampling hole. This paper describes the architecture, performances and return of experience of the long distance fire detection system developed and installed to secure the CERN Super Proton Synchrotron tunnels.

Keywords: air sampling, fire detection, long distance, radioactive areas

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Authors: Ankita Jadon, Sidi Souvi, Nathalie Girault, Denis Petitprez

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The interaction of fission products (FP) with sodium carbonate (Na₂CO₃) aerosols is of a high safety concern because of their potential role in the radiological source term mitigation by FP trapping. In a sodium-cooled fast nuclear reactor (SFR) experiencing a severe accident, sodium (Na) aerosols can be formed after the ejection of the liquid Na coolant inside the containment. The surface interactions between these aerosols and different FP species have been investigated using ab-initio, density functional theory (DFT) calculations using Vienna ab-initio simulation package (VASP). In addition, an improved thermodynamic model has been proposed to treat DFT-VASP calculated energies to extrapolate them to temperatures and pressures of interest in our study. A combined experimental and theoretical chemistry study has been carried out to have both atomistic and macroscopic understanding of the chemical processes; the theoretical chemistry part of this approach is presented in this paper. The Perdew, Burke, and Ernzerhof functional were applied in combination with Grimme’s van der Waals correction to compute exchange-correlational energy at 0 K. Seven different surface cleavages were studied of Ƴ-Na₂CO₃ phase (stable at 603.15 K), it was found that for defect-free surfaces, the (001) facet is the most stable. Furthermore, calculations were performed to study surface defects and reconstructions on the ideal surface. All the studied surface defects were found to be less stable than the ideal surface. More than one adsorbate-ligand configurations were found to be stable confirming that FP vapors could be trapped on various adsorption sites. The calculated adsorption energies (Eads, eV) for the three most stable adsorption sites for I₂ are -1.33, -1.088, and -1.085. Moreover, the adsorption of the first molecule of I₂ changes the surface in a way which would favor stronger adsorption of a second molecule of I2 (Eads, eV = -1.261). For HI adsorption, the most favored reactions have the following Eads (eV) -1.982, -1.790, -1.683 implying that HI would be more reactive than I₂. In addition to FP species, adsorption of H₂O was also studied as the hydrated surface can have different reactivity than the bare surface. One thermodynamically favored site for H₂O adsorption was found with an Eads, eV of -0.754. Finally, the calculations of hydrated surfaces of Na₂CO₃ show that a layer of water adsorbed on the surface significantly reduces its affinity for iodine (Eads, eV = -1.066). According to the thermodynamic model built, the required partial pressure at 373 K to have adsorption of the first layer of iodine is 4.57×10⁻⁴ bar. The second layer will be adsorbed at partial pressures higher than 8.56×10⁻⁶ bar; a layer of water on the surface will increase these pressure almost ten folds to 3.71×10⁻³ bar. The surface interacts with elemental Cs with an Eads (eV) of -1.60, while interacts even strongly with CsI with an Eads (eV) of -2.39. More results on the interactions between Na₂CO₃ (001) and cesium-based FP will also be presented in this paper.

Keywords: iodine uptake, sodium carbonate surface, sodium-cooled fast nuclear reactor, DFT calculations, fission products

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Authors: Ki Hyun Park, Hye Min Park, Jeong Ho Kim, Chan Jong Park, Koan Sik Joo

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In the modern days, the use of radioactive substances is on the rise in the areas like chemical weaponry, industrial usage, and power plants. Although there are various technologies available to detect and monitor radioactive substances in the air, the technologies to detect underwater radioactive substances are scarce. In this study, computer simulation of the underwater detection system measuring beta-ray, a radioactive substance, has been done through MCNP. CaF₂, YAP(Ce) and YAG(Ce) have been used in the computer simulation to detect beta-ray as scintillator. Also, the source used in the computer simulation is Sr-90 and Y-90, both of them emitting only pure beta-ray. The distance between the source and the detector was shifted from 1mm to 10mm by 1 mm in the computer simulation. The result indicated that Sr-90 was impossible to measure below 1 mm since its emission energy is low while Y-90 was able to be measured up to 10mm underwater. In addition, the detector designed with CaF₂ had the highest efficiency among 3 scintillators used in the computer simulation. Since it was possible to verify the detectable range and the detection efficiency according to modeling through MCNP simulation, it is expected that such result will reduce the time and cost in building the actual beta-ray detector and evaluating its performances, thereby contributing the research and development.

Keywords: Beta-ray, CaF₂, detector, MCNP simulation, scintillator

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Authors: Longkui Zhu, Zhengcao Li

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High temperature gas-cooled reactors (HTGRs) are considered as one of the next-generation advanced nuclear reactors, in which porous nuclear graphite is used as neutron moderators, reflectors, structure materials, and cooled by inert helium. Radioactive tritium and ¹⁴C are generated in terms of reactions of thermal neutrons and ⁶Li, ¹⁴N, ¹⁰B impurely within nuclear graphite and the coolant during HTGRs operation. Currently, hydrogen and nitrogen diffusion behavior together with nuclear graphite microstructure evolution were investigated to minimize the radioactive waste release, using thermogravimetric analysis, X-ray computed tomography, the BET and mercury standard porosimetry methods. It is found that the peak value of graphite weight loss emerged at 573-673 K owing to nitrogen diffusion from graphite pores to outside when the system was subjected to vacuum. Macropore volume became larger while porosity for mesopores was smaller with temperature ranging from ambient temperature to 1073 K, which was primarily induced by coalescence of the subscale pores. It is suggested that the porous nuclear graphite should be first subjected to vacuum at 573-673 K to minimize the nitrogen and the radioactive 14°C before operation in HTGRs. Then, results on hydrogen diffusion show that the diffusible hydrogen and tritium could permeate into the coolant with diffusion coefficients of > 0.5 × 10⁻⁴ cm²·s⁻¹ at 50 bar. As a consequence, the freshly-generated diffusible tritium could release quickly to outside once formed, and an effective approach for minimizing the amount of radioactive tritium is to make the impurity contents extremely low in nuclear graphite and the coolant. Besides, both two- and three-dimensional observations indicate that macro and mesopore volume along with total porosity decreased with temperature at 50 bar on account of synergistic effects of applied compression strain, sharpened pore morphology, and non-uniform temperature distribution.

Keywords: advanced high temperature gas-cooled reactor, hydrogen and nitrogen diffusion, microstructure evolution, nuclear graphite, radioactive waste management

Procedia PDF Downloads 285

Authors: Shiyi He, Yan Xia, Xiaoping Ouyang, Liang Chen, Zhongbing Zhang, Jinlu Ruan

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As the key system of the spacecraft, various propelling system have been developing rapidly, including ion thrust, laser thrust, solar sail and other micro-thrusters. However, there still are some shortages in these systems. The ion thruster requires the high-voltage or magnetic field to accelerate, resulting in extra system, heavy quantity and large volume. The laser thrust now is mostly ground-based and providing pulse thrust, restraint by the station distribution and the capacity of laser. The thrust direction of solar sail is limited to its relative position with the Sun, so it is hard to propel toward the Sun or adjust in the shadow.In this paper, a novel nuclear thruster based on alpha decay and spontaneous fission is proposed and the principle of this radiation-thrust with alpha particle has been expounded. Radioactive materials with different released energy, such as 210Po with 5.4MeV and 238Pu with 5.29MeV, attached to a metal film will provides various thrust among 0.02-5uN/cm2. With this repulsive force, radiation is able to be a power source. With the advantages of low system quantity, high accuracy and long active time, the radiation thrust is promising in the field of space debris removal, orbit control of nano-satellite array and deep space exploration. To do further study, a formula lead to the amplitude and direction of thrust by the released energy and decay coefficient is set up. With the initial formula, the alpha radiation elements with the half life period longer than a hundred days are calculated and listed. As the alpha particles emit continuously, the residual charge in metal film grows and affects the emitting energy distribution of alpha particles. With the residual charge or extra electromagnetic field, the emitting of alpha particles performs differently and is analyzed in this paper. Furthermore, three more complex situations are discussed. Radiation element generating alpha particles with several energies in different intensity, mixture of various radiation elements, and cascaded alpha decay are studied respectively. In combined way, it is more efficient and flexible to adjust the thrust amplitude. The propelling model of the spontaneous fission is similar with the one of alpha decay, which has a more complex angular distribution. A new quasi-sphere space propelling system based on the radiation-thrust has been introduced, as well as the collecting and processing system of excess charge and reaction heat. The energy and spatial angular distribution of emitting alpha particles on unit area and certain propelling system have been studied. As the alpha particles are easily losing energy and self-absorb, the distribution is not the simple stacking of each nuclide. With the change of the amplitude and angel of radiation-thrust, orbital variation strategy on space debris removal is shown and optimized.

Keywords: alpha decay, angular distribution, emitting energy, orbital variation, radiation-thruster

Procedia PDF Downloads 163

Authors: Edward Holupka, John Rossman, Tye Morancy, Joseph Aronovitz, Irving Kaplan

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A common modality for the treatment of early stage prostate cancer is the implantation of radioactive seeds directly into the prostate. The radioactive seeds are positioned inside the prostate to achieve optimal radiation dose coverage to the prostate. These radioactive seeds are positioned inside the prostate using Transrectal ultrasound imaging. Once all of the planned seeds have been implanted, two dimensional transaxial transrectal ultrasound images separated by 2 mm are obtained through out the prostate, beginning at the base of the prostate up to and including the apex. A common deep neural network, called DetectNet was trained to automatically determine the position of the implanted radioactive seeds within the prostate under ultrasound imaging. The results of the training using 950 training ultrasound images and 90 validation ultrasound images. The commonly used metrics for successful training were used to evaluate the efficacy and accuracy of the trained deep neural network and resulted in an loss_bbox (train) = 0.00, loss_coverage (train) = 1.89e-8, loss_bbox (validation) = 11.84, loss_coverage (validation) = 9.70, mAP (validation) = 66.87%, precision (validation) = 81.07%, and a recall (validation) = 82.29%, where train and validation refers to the training image set and validation refers to the validation training set. On the hardware platform used, the training expended 12.8 seconds per epoch. The network was trained for over 10,000 epochs. In addition, the seed locations as determined by the Deep Neural Network were compared to the seed locations as determined by a commercial software based on a one to three months after implant CT. The Deep Learning approach was within \strikeout off\uuline off\uwave off2.29\uuline default\uwave default mm of the seed locations determined by the commercial software. The Deep Learning approach to the determination of radioactive seed locations is robust, accurate, and fast and well within spatial agreement with the gold standard of CT determined seed coordinates.

Keywords: prostate, deep neural network, seed implant, ultrasound

Procedia PDF Downloads 163

Authors: Claudiu Mazilu, Ion Robu, Radu Deju

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At the end of 2011 worldwide there were 124 power reactors shut down, from which: 16 fully decommissioned, 50 power reactors in a decommissioning process, 49 reactors in “safe enclosure mode”, 3 reactors “entombed”, for other 6 reactors it was not yet have specified the decommissioning strategy. The concrete radioactive waste that will be generated from dismantled structures of VVR-S nuclear research reactor from Magurele (e.g.: biological shield of the reactor core and hot cells) represents an estimated amount of about 70 tons. Until now the solid low activity radioactive waste (LLW) was pre-placed in containers and cementation with mortar made from cement and natural fine aggregates, providing a fill ratio of the container of approximately 50 vol. % for concrete. In this paper is presented an innovative technology in which radioactive concrete is crushed and the mortar made from recycled radioactive sand, cement, water and superplasticizer agent is poured in container with radioactive rubble (that is pre-placed in container) for cimentation. Is achieved a radioactive waste package in which the degree of filling of radioactive waste increases substantially. The tests were carried out on non-radioactive material because the radioactive concrete was not available in a good time. Waste concrete with maximum size of 350 mm were crushed in the first stage with a Liebhher type jaw crusher, adjusted to nominal size of 50 mm. Crushed concrete less than 50 mm was sieved in order to obtain useful sort for preplacement, 10 to 50 mm. The rest of the screening > 50 mm obtained from primary crushing of concrete was crushed in the second stage, with different working principles crushers at size < 2.5 mm, in order to produce recycled fine aggregate (sand) for the filler mortar and which fulfills the technical specifications proposed: –jaw crusher, Retsch type, model BB 100; –hammer crusher, Buffalo Shuttle model WA-12-H; presented a series of characteristics of recycled concrete aggregates by predefined class (the granulosity, the granule shape, the absorption of water, behavior to the Los Angeles test, the content of attached mortar etc.), most in comparison with characteristics of natural aggregates. Various mortar recipes were used in order to identify those that meet the proposed specification (flow-rate: 16-50s, no bleeding, min. 30N/mm2 compressive strength of the mortar after 28 days, the proportion of recycled sand used in mortar: min. 900kg/m3) and allow obtaining of the highest fill ratio for mortar. In order to optimize the mortars following compositional factors were varied: aggregate nature, water/cement (W/C) ratio, sand/cement (S/C) ratio, nature and proportion of additive. To confirm the results obtained on a small scale, it made an attempt to fill the mortar in a container that simulates the final storage drums. Was measured the mortar fill ratio (98.9%) compared with the results of laboratory tests and targets set out in the proposed specification. Although fill ratio obtained on the mock-up is lower by 0.8 vol. % compared to that obtained in the laboratory tests (99.7%), the result meets the specification criteria.

Keywords: characteristics, radioactive recycled concrete aggregate, mortars, fill ratio

Procedia PDF Downloads 167

Authors: B. Sadepovich Maikanov, Z. Shabanbayevich Adilbekov, R. Husainovna Mustafina, L. Tyulegenovna Auteleyeva

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This study involves detailed information about contaminants of honey in the Republic of Kazakhstan. The requirements of the technical regulation ‘Requirements to safety of honey and bee products’ and GOST 19792-2001 were taken into account in this research. Contamination of honey by antibiotics wqs determined by the IEA (immune-enzyme analysis), Ridder analyzer and Tecna produced test systems. Voltammetry (TaLab device) was used to define contamination by salts of heavy metals and gamma-beta spectrometry, ‘Progress BG’ system, with preliminary ashing of the sample of honey was used to define radioactive contamination. This article pointed out that residues of chloramphenicol were detected in 24% of investigated products, in 22% of them –streptomycin, in 7.3% - sulfanilamide, in 2.4% - tylosin, and in 12% - combined contamination was noted. Geographically, the greatest degree of contamination of honey with antibiotics occurs in the Northern Kazakhstan – 54.4%, and Southern Kazakhstan - 50%, and the lowest in Central and Eastern Kazakhstan with 30% and 25%, respectively. Generally, pollution by heavy metals is within acceptable limits, but the contamination from lead is highest in the Akmola region. The level of radioactive cesium and strontium is also within acceptable concentrations. The highest radioactivity in terms of cesium was observed in the East Kazakhstan region - 49.00±10 Bq/kg, in Akmola, North Kazakhstan and Almaty - 12.00±5, 11.05±3 and 19.0±8 Bq/kg, respectively, while the norm is 100 Bq/kg. In terms of strontium, the radioactivity in the East Kazakhstan region is 25.03±15 Bq/kg, while in Akmola, North Kazakhstan and Almaty regions it is 12.00±3, 10.2±4 and 1.0±2 Bq/kg, respectively, with the norm of 80 Bq/kg. This accumulation is mainly associated with the environmental degradation, feeding and treating of bees. Moreover, in the process of collecting nectar, external substances can penetrate honey. Overall, this research determines factors and reasons of honey contamination.

Keywords: antibiotics, contamination of honey, honey, radionuclides

Procedia PDF Downloads 185

Authors: M. I. A. Adziz, J. Sharib Sarip, M. T. Ishak, D. N. A. Tugi

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Radiation exposure to humans and the environment is caused by natural radioactive material sources. Given that exposure to people and communities can occur through several pathways, it is necessary to pay attention to the increase in naturally radioactive material, particularly in the soil. Continuous research and monitoring on the distribution and determination of these natural radionuclides' activity as a guide and reference are beneficial, especially in an accidental exposure. Surface soil/sediment samples from several locations identified around the Sembrong catchment area were taken for the study. After 30 days of secular equilibrium with their daughters, the activity concentrations of the naturally occurring radioactive material (NORM) members, i.e. ²²⁶Ra, ²²⁸Ra, ²³⁸U, ²³²Th, and ⁴⁰K, were measured using high purity germanium (HPGe) gamma spectrometer. The results obtained showed that the radioactivity concentration of ²³⁸U ranged between 17.13 - 30.13 Bq/kg, ²³²Th ranged between 22.90 - 40.05 Bq/kg, ²²⁶Ra ranged between 19.19 - 32.10 Bq/kg, ²²⁸Ra ranged between 21.08 - 39.11 Bq/kg and ⁴⁰K ranged between 9.22 - 51.07 Bq/kg with average values of 20.98 Bq/kg, 27.39 Bq/kg, 23.55 Bq/kg, 26.93 Bq/kg and 23.55 Bq/kg respectively. The values obtained from this study were low or equivalent to previously reported in previous studies. It was also found that the mean/mean values obtained for the four parameters of the Radiation Hazard Index, namely radium equivalent activity (Raeq), external dose rate (D), annual effective dose and external hazard index (Hₑₓ), were 65.40 Bq/kg, 29.33 nGy/h, 19.18 ¹⁰⁻⁶Sv and 0.19 respectively. These obtained values are low compared to the world average values and the values of globally applied standards. Comparison with previous studies (dry season) also found that the values for all four parameters were low and equivalent. This indicates the level of radiation hazard in the area around the study is safe for the public.

Keywords: catchment area, gamma spectrometry, naturally occurring radioactive material (NORM), soil

Procedia PDF Downloads 67

Authors: Paul Essel, Eric T. Glover, Gustav Gbeddy, Yaw Adjei-Kyereme, Abdallah M. A. Dawood, Evans M. Ameho, Emmanuel A. Aberikae

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Radioactive materials mainly in the form of Sealed Radioactive Sources are being used in various sectors (medicine, agriculture, industry, research, and teaching) for the socio-economic development of Ghana. The use of these beneficial radioactive materials has resulted in an inventory of Disused Sealed Radioactive Sources (DSRS) in storage. Most of the DSRS are legacy/historic sources which cannot be returned to their manufacturer or country of origin. Though small in volume, DSRS can be intensively radioactive and create a significant safety and security liability. They need to be managed in a safe and secure manner in accordance with the fundamental safety objective. The Radioactive Waste Management Center (RWMC) of the Ghana Atomic Energy Commission (GAEC) is currently storing a significant volume of DSRS. The initial activities of the DSRS range from 7.4E+5 Bq to 6.85E+14 Bq. If not managed properly, such DSRS can represent a potential hazard to human health and the environment. Storage is an important interim step, especially for DSRS containing very short-lived radionuclides, which can decay to exemption levels within a few years. Long-term storage, however, is considered an unsustainable option for DSRS with long half-lives hence the need for a disposal facility. The GAEC intends to use the International Atomic Energy Agency’s (IAEA’s) Borehole Disposal System (BDS) to provide a safe, secure, and cost-effective disposal option to dispose of its DSRS in storage. The proposed site for implementation of the BDS is on the GAEC premises at Kwabenya. The site has been characterized to gain a general understanding in terms of its regional setting, its past evolution and likely future natural evolution over the assessment time frame. Due to the long half-lives of some of the radionuclides to be disposed of (Ra-226 with half-life of 1600 years), the engineered barriers of the system must be robust to contain these radionuclides for this long period before they decay to harmless levels. There is the need to assess the safety related functions of the engineered barriers of this disposal system.

Keywords: radionuclides, disposal, radioactive waste, engineered barrier

Procedia PDF Downloads 18

Authors: Hugo Lefeuvre, Jerôme Donnard, Michael Descostes, Sophie Billon, Samuel Duval, Tugdual Oger, Herve Toubon, Paul Sardini

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Spectroscopic autoradiography is a method of interest for geological sample analysis. Indeed, researchers may face different issues such as radioelement identification and quantification in the field of environmental studies. Imaging gaseous ionization detectors find their place in geosciences for conducting specific measurements of radioactivity to improve the monitoring of natural processes using naturally-occurring radioactive tracers, but also for the nuclear industry linked to the mining sector. In geological samples, the location and identification of the radioactive-bearing minerals at the thin-section scale remains a major challenge as the detection limit of the usual elementary microprobe techniques is far higher than the concentration of most of the natural radioactive decay products. The spatial distribution of each decay product in the case of uranium in a geomaterial is interesting for relating radionuclides concentration to the mineralogy. The present study aims to provide spectroscopic autoradiography analysis method for measuring the initial energy of alpha particles with a parallel ionization multiplier gaseous detector. The analysis method has been developed thanks to Geant4 modelling of the detector. The track of alpha particles recorded in the gas detector allow the simultaneous measurement of the initial point of emission and the reconstruction of the initial particle energy by a selection based on the linear energy distribution. This spectroscopic autoradiography method was successfully used to reproduce the alpha spectra from a 238U decay chain on a geological sample at the thin-section scale. The characteristics of this measurement are an energy spectrum resolution of 17.2% (FWHM) at 4647 keV and a spatial resolution of at least 50 µm. Even if the efficiency of energy spectrum reconstruction is low (4.4%) compared to the efficiency of a simple autoradiograph (50%), this novel measurement approach offers the opportunity to select areas on an autoradiograph to perform an energy spectrum analysis within that area. This opens up possibilities for the detailed analysis of heterogeneous geological samples containing natural alpha emitters such as uranium-238 and radium-226. This measurement will allow the study of the spatial distribution of uranium and its descendants in geo-materials by coupling scanning electron microscope characterizations. The direct application of this dual modality (energy-position) of analysis will be the subject of future developments. The measurement of the radioactive equilibrium state of heterogeneous geological structures, and the quantitative mapping of 226Ra radioactivity are now being actively studied.

Keywords: alpha spectroscopy, digital autoradiography, mining activities, natural decay products

Procedia PDF Downloads 116

Authors: Kulwinder Singh Mann, Manmohan Singh Heer, Asha Rani

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The safe storage of radioactive materials has become an important issue. Nuclear engineering necessitates the safe handling of radioactive materials emitting high energy gamma-rays. Hazards involved in handling radioactive materials insist suitable shielded enclosures. With overgrowing use of nuclear energy for meeting the increasing demand of power, there is a need to investigate the shielding behavior of cost effective shielded enclosure (CESE) made from clay-bricks (CB) and fire-bricks (FB). In comparison to the lead-bricks (conventional-shielding), the CESE are the preferred choice in nuclear waste management. The objective behind the present investigation is to evaluate the double layered transmission exposure buildup factors (DLEBF) for gamma-rays for CESE in energy range 0.5-3MeV. For necessary computations of shielding parameters, using existing huge data regarding gamma-rays interaction parameters of all periodic table elements, two computer programs (GRIC-toolkit and BUF-toolkit) have been designed. It has been found that two-layered slabs show effective shielding for gamma-rays in orientation CB followed by FB than the reverse. It has been concluded that the arrangement, FB followed by CB reduces the leakage of scattered gamma-rays from the radioactive source.

Keywords: buildup factor, clay bricks, fire bricks, nuclear wastage management, radiation protective double layered slabs

Procedia PDF Downloads 365