Search results for: radioactive waste

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: 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: 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: Sou Watanabe, Hiromichi Ogi, Atsuhiro Shibata, Kazunori Nomura

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As a part of STRAD project conducted by JAEA, condensation of radioactive liquid waste containing various chemical compounds using reverse osmosis (RO) membrane filter was examined for efficient and safety treatment of the liquid wastes accumulated inside hot laboratories. NH₄⁺ ion in the feed solution was successfully concentrated, and NH₄⁺ ion involved in the effluents became lower than target value; 100 ppm. Solidification of simulated aqueous and organic liquid wastes was also tested. Those liquids were successfully solidified by adding cement or coagulants. Nevertheless, optimization in materials for confinement of chemicals is required for long time storage of the final solidified wastes.

Keywords: condensation, radioactive liquid waste, solidification, STRAD project

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Authors: R. Deju, M. Mincu, D. Gurau

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During the interim storage or final disposal of low level waste, migration/diffusion of radionuclides can occur when the waste comes in contact with water. The long-term leaching behaviour into surrounding fluid (demineralized water) of ¹³⁷Cs, ⁶⁰Co and ¹⁵²Eu radionuclides, artificially incorporated in mortar matrices made from natural aggregates (river sand) and recycled radioactive concrete was studied. Results presented in this work are obtained in two years of mortar testing and will be used for the safety increasing in the storage of low level radioactive waste. The study involved the influence of curing time, type and size distribution of the aggregates on leaching behaviour. The mortar samples were immersed in distilled water for 30 days. The leached activity of the mortar samples was measured on samples from the immersing water and analyzed through a gamma-ray spectrometry method using an HPGe detector with a GESPECOR code for efficiency evaluation. The long-term leaching behaviour of the radionuclides was evaluated from the leaching data calculating the apparent diffusion coefficient.

Keywords: gamma spectrometry, leaching behavior, reuse and recycling of radioactive concrete, waste management

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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: 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: Kyomin Lee, Joohee Kim, Sangho Kang

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The commercial nuclear power reactor in South Korea, Kori Unit 1, which was a 587 MWe pressurized water reactor that started operation since 1978, was permanently shut down in June 2017 without an additional operating license extension. The Kori 1 Unit is scheduled to become the nuclear power unit to enter the decommissioning phase. In this study, the preliminary evaluation of the decommissioning wastes for the Kori Unit 1 was performed based on the following series of process: firstly, the plant inventory is investigated based on various documents (i.e., equipment/ component list, construction records, general arrangement drawings). Secondly, the radiological conditions of systems, structures and components (SSCs) are established to estimate the amount of radioactive waste by waste classification. Third, the waste management strategies for Kori Unit 1 including waste packaging are established. Forth, selection of the proper decontamination and dismantling (D&D) technologies is made considering the various factors. Finally, the amount of decommissioning waste by classification for Kori 1 is estimated using the DeCAT program, which was developed by KEPCO-E&C for a decommissioning cost estimation. The preliminary evaluation results have shown that the expected amounts of decommissioning wastes were less than about 2% and 8% of the total wastes generated (i.e., sum of clean wastes and radwastes) before/after waste processing, respectively, and it was found that the majority of contaminated material was carbon or alloy steel and stainless steel. In addition, within the range of availability of information, the results of the evaluation were compared with the results from the various decommissioning experiences data or international/national decommissioning study. The comparison results have shown that the radioactive waste amount from Kori Unit 1 decommissioning were much less than those from the plants decommissioned in U.S. and were comparable to those from the plants in Europe. This result comes from the difference of disposal cost and clearance criteria (i.e., free release level) between U.S. and non-U.S. The preliminary evaluation performed using the methodology established in this study will be useful as a important information in establishing the decommissioning planning for the decommissioning schedule and waste management strategy establishment including the transportation, packaging, handling, and disposal of radioactive wastes.

Keywords: characterization, classification, decommissioning, decontamination and dismantling, Kori 1, radioactive waste

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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

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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

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

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Chemical Processing Facility of Japan Atomic Energy Agency is a basic research field for advanced back-end technology developments with using actual high-level radioactive materials such as irradiated fuels from the fast reactor, high-level liquid waste from reprocessing plant. In the nature of a research facility, various kinds of chemical reagents have been offered for fundamental tests. Most of them were treated properly and stored in the liquid waste vessel equipped in the facility, but some were not treated and remained at the experimental space as a kind of legacy waste. It is required to treat the waste in safety. On the other hand, we formulated the Medium- and Long-Term Management Plan of Japan Atomic Energy Agency Facilities. This comprehensive plan considers Chemical Processing Facility as one of the facilities to be decommissioned. Even if the plan is executed, treatment of the “legacy” waste beforehand must be a necessary step for decommissioning operation. Under this circumstance, we launched a collaborative research project called the STRAD project, which stands for Systematic Treatment of Radioactive liquid waste for Decommissioning, in order to develop the treatment processes for wastes of the nuclear research facility. In this project, decomposition methods of chemicals causing a troublesome phenomenon such as corrosion and explosion have been developed and there is a prospect of their decomposition in the facility by simple method. And solidification of aqueous or organic liquid wastes after the decomposition has been studied by adding cement or coagulants. Furthermore, we treated experimental tools of various materials with making an effort to stabilize and to compact them before the package into the waste container. It is expected to decrease the number of transportation of the solid waste and widen the operation space. Some achievements of these studies will be shown in this paper. The project is expected to contribute beneficial waste management outcome that can be shared world widely.

Keywords: chemical processing facility, medium- and long-term management plan of JAEA facilities, STRAD project, treatment of radioactive waste

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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: 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

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Authors: Eve-Andree Laramee

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The aesthetics and ethics of the Nuclear Anthropocene are explored through artists responses to the impact of radioactive materials on ecological systems, global issues, energy policies and ourselves. This presentation tracks and reveals the invisible traces of the nuclear weapons complex and the nuclear energy industry, in relation to environmental justice. Radioactive pollution transgresses international borders, boundaries between land and water, contaminating ecological systems. Radioactive waste is never disposed of; it is dispositioned, placed out of sight and out of mind. These materials leave behind an invisible toxic legacy lasting millions of years. As we are learning post-Fukushima, when climate change occurs and vulnerability spectrums shift, nuclear sites and the life forms surrounding them are at increased risk. By visualizing this contamination through art installations, videos, and social-sculpture interventions, information is shared with the public, raising awareness, and activating community participation in remediation and nonproliferation efforts. The emerging Ecological Art genre proposes paradigms sustainable with the life forms and resources of our planet. It is comprised of artists, scientists, philosophers and activists devoted to these. EcoArt is distinguished by a focus on systems and interrelationships within our environment: the ecological, geographic, political, biological and cultural. This presentation will cover artworks addressing the recent Fukushima meltdowns, weapons proliferation, climate change, radioactive waste disposal and environmental justice. Possibilities for art-and-science collaborations will be discussed as projects that sharpen our ethics and politics in our behaviors and social interactions. The presentation will consist of a PowerPoint talk (paper presentation) accompanied by images and video clips.

Keywords: art, ecology, environment, anthropocene, nuclear

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Authors: Kazuyuki Takase, Yoshihisa Hiraki, Gaku Takase, Isamu Kudo

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One of the most important issue is to ensure the safety of long-term waste storage containers in which fuel debris and radioactive materials are accumulated. In this case, hydrogen generated by water decomposition by radiation is accumulated in the container for a long period of time, so it is necessary to reduce the concentration of hydrogen in the container. In addition, a condition that any power supplies from the outside of the container are unnecessary is requested. Then, radioactive waste storage containers with the passive autocatalytic recombiner (PAR) would be effective. The radioactive waste storage container with PAR was used for moving the fuel debris of the Three Mile Island Unit 2 to the storage location. However, the effect of PAR is not described in detail. Moreover, the reduction of hydrogen concentration during the long-term storage period was performed by the venting system, which was installed on the top of the container. Therefore, development of a long-term storage container with PAR was started with the aim of safely storing fuel debris picked up at the Fukushima Daiichi Nuclear Power Plant for a long period of time. A fundamental experiment for reducing the concentration of hydrogen which generates in a nuclear waste long-term storage container was carried out using a small-scale container with PAR. Moreover, the circulation flow behavior of hydrogen in the small-scale container resulting from the natural convection by the decay heat was clarified. In addition, preliminary numerical analyses were performed to predict the experimental results regarding the circulation flow behavior and the reduction of hydrogen concentration in the small-scale container. From the results of the present study, the validity of the container with PAR was experimentally confirmed on the reduction of hydrogen concentration. In addition, it was predicted numerically that the circulation flow behavior of hydrogen in the small-scale container is blocked by steam which generates by chemical reaction of hydrogen and oxygen.

Keywords: hydrogen behavior, reduction of concentration, long-term storage container, small-scale, PAR, experiment, analysis

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Authors: Babanyara Y. Y., Ibrahim D. B., Garba T., Bogoro A. G., Abubakar, M. Y.

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Medical care is vital for our life, health, and well-being. However, the waste generated from medical activities can be hazardous, toxic, and even lethal because of their high potential for diseases transmission. The hazardous and toxic parts of waste from healthcare establishments comprising infectious, medical, and radioactive material as well as sharps constitute a grave risks to mankind and the environment, if these are not properly treated/disposed or are allowed to be mixed with other municipal waste. In Nigeria, practical information on this aspect is inadequate and research on the public health implications of poor management of medical wastes is few and limited in scope. Findings drawn from Literature particularly in the third world countries highlights financial problems, lack of awareness of risks involved in MWM, lack of appropriate legislation and lack of specialized MWM staff. The paper recommends how MWM practices can be improved in medical facilities.

Keywords: environmental pollution, infectious, management, medical waste, public health

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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: 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

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Authors: Silvia Carla Da Silva André, Angela Maria Magosso Takayanagui

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Introduction: The Medical Waste (MW) are responsible per 2% of total waste generated for a city and has merited attention due the risks that offers to the public health and environment, representing an important aspect in waste management. In Brazil, the Resolution 306/04 of the National Health Surveillance Agency classifies the MW into 5 groups as follows: Group A (GA) biological, Group B (GB) chemical, Group C (GC) radioactive waste, Group D (GD) common, and Group E (GE) sharps. Objective: This study aimed to determine the amount of waste generated in hospitals of Ribeirão Preto, São Paulo, Brazil. Material and Methods: This is a field research, exploratory, using quantitative variables. The survey was conducted in 11 hospitals in Ribeirão Preto, located in the State of São Paulo, Brazil. It is noted that the study sample included general hospitals, skilled, university, maternity, and psychiatric; public, private, and philanthropic; and large, medium, and small. To quantify the MW, the weighing of the waste was held for six days, following methodology adapted from PAHO. Data were analyzed using descriptive statistics, determining the average global generation of MW and for each group. This research was carried out after approval by the Ethics in Research of the University of São Paulo. Thus, in order to comply with the ethical principles of research, to present the results hospitals were numbered from 1 to 11. Results: The data revealed a greater generation of biological waste among teaching hospitals, which can be justified by the use of materials for the realization of techniques.

Keywords: environmental health, management of medical waste, medical waste, public health

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Authors: J. W. Choi, S. Y. Cho, H. J. Lee, W. Z. Oh, S. J. Choi

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Ionic liquids (ILs), which is alternative to conventional organic solvent, were used for extraction of Cs ions. ILs, as useful environment friendly green solvents, have been recently applied as replacement for traditional volatile organic compounds (VOCs) in liquid/liquid extraction of heavy metal ions as well as organic and inorganic species and pollutants. Thus, Ionic liquids were used for extraction of Cs ions from the liquid radioactive waste. In most cases, Cs ions present in radioactive wastes in very low concentration, approximately less than 1ppm. Therefore, unlike established extraction system the required amount of ILs as extractant is comparatively very small. This extraction method involves cation exchange mechanism in which Cs ion transfers to the organic phase and binds to one crown ether by chelation in exchange of single ILs cation, IL_cation+, transfer to the aqueous phase. In this extraction system showed solid-liquid separation in which the Ionic liquid 1-ethyl-3-methylimidazolium bis(trifluoromethylsulfonly)imide (C2mimTf2N) and the crown ether Dicyclohexano-18-crown-6 (DCH18C6) both were used here in very little amount as solvent and as extractant, respectively. 30 mM of CsNO3 was used as simulated waste solution cesium ions. Generally, in liquid-liquid extraction, the molar ratio of CE:Cs+:ILs was 1:5~10:>100, while our applied molar ratio of CE:Cs+:ILs was 1:2:1~10. The quantity of CE and Cs ions were fixed to 0.6 and 1.2 mmol, respectively. The phenomenon of precipitation showed two kinds of separation: solid-liquid separation in the ratio of 1:2:1 and 1:2:2; solid-liquid-liquid separation (3 phase) in the ratio of 1:2:5 and 1:2:10. In the last system, 3 phases were precipitate-ionic liquids-aqueous. The precipitate was verified to consist of Cs+, DCH18C6, Tf2N- based on the cation exchange mechanism. We analyzed precipitate using scanning electron microscopy with X-ray microanalysis (SEM-EDS), an elemental analyser, Fourier transform infrared spectroscopy (FT-IR) and differential scanning calorimetry (DSC). The experimental results showed an easy extraction method and confirmed the composition of solid precipitate. We also obtained information that complex formation ratio of Cs+ to DCH18C6 is 0.88:1 regardless of C2mimTf2N quantities.

Keywords: extraction, precipitation, solid-liquid seperation, ionic liquid, precipitate

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Authors: Fatima Zahra Falah, Touria El. Ghailassi, Samia Yousfi, Ahmed Moussaif, Hasna Hamdane, Mouna Latifa Bouamrani

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The management and treatment of radioactive wastewater pose significant challenges to environmental safety and public health. This study presents an innovative approach to optimizing radioactive wastewater treatment using a novel biosorbent: chitosan-argan nutshell beads. By employing Response Surface Methodology (RSM), we aimed to determine the optimal conditions for maximum removal efficiency of radioactive contaminants. Chitosan, a biodegradable and non-toxic biopolymer, was combined with argan nutshell powder to create composite beads. The argan nutshell, a waste product from argan oil production, provides additional adsorption sites and mechanical stability to the biosorbent. The beads were characterized using Fourier Transform Infrared Spectroscopy (FTIR), Scanning Electron Microscopy (SEM), and X-ray Diffraction (XRD) to confirm their structure and composition. A three-factor, three-level Box-Behnken design was utilized to investigate the effects of pH (3-9), contact time (30-150 minutes), and adsorbent dosage (0.5-2.5 g/L) on the removal efficiency of radioactive isotopes, primarily focusing on cesium-137. Batch adsorption experiments were conducted using synthetic radioactive wastewater with known concentrations of these isotopes. The RSM analysis revealed that all three factors significantly influenced the adsorption process. A quadratic model was developed to describe the relationship between the factors and the removal efficiency. The model's adequacy was confirmed through analysis of variance (ANOVA) and various diagnostic plots. Optimal conditions for maximum removal efficiency were pH 6.8, a contact time of 120 minutes, and an adsorbent dosage of 0.8 g/L. Under these conditions, the experimental removal efficiency for cesium-137 was 94.7%, closely matching the model's predictions. Adsorption isotherms and kinetics were also investigated to elucidate the mechanism of the process. The Langmuir isotherm and pseudo-second-order kinetic model best described the adsorption behavior, indicating a monolayer adsorption process on a homogeneous surface. This study demonstrates the potential of chitosan-argan nutshell beads as an effective and sustainable biosorbent for radioactive wastewater treatment. The use of RSM allowed for the efficient optimization of the process parameters, potentially reducing the time and resources required for large-scale implementation. Future work will focus on testing the biosorbent's performance with real radioactive wastewater samples and investigating its regeneration and reusability for long-term applications.

Keywords: adsorption, argan nutshell, beads, chitosan, mechanism, optimization, radioactive wastewater, response surface methodology

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Authors: Joseph Lillington, Tom Gout, Mike Harrison, Ian Farnan

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The vitrification of high-level nuclear waste within borosilicate glass and its incorporation within a multi-barrier repository deep underground is widely accepted as the preferred disposal method. However, for this to happen, any safety case will require validation that the initially localized radionuclides will not be considerably released into the near/far-field. Therefore, accurate mechanistic models are necessary to predict glass dissolution, and these should be robust to a variety of incorporated waste species and leaching test conditions, particularly given substantial variations across international waste-streams. Here, machine learning is used to predict glass material properties (viscosity, density) and glass leaching model parameters from large-scale industrial data. A variety of different machine learning algorithms have been compared to assess performance. Density was predicted solely from composition, whereas viscosity additionally considered temperature. To predict suitable glass leaching model parameters, a large simulated dataset was created by coupling MATLAB and the chemical reactive-transport code HYTEC, considering the state-of-the-art GRAAL model (glass reactivity in allowance of the alteration layer). The trained models were then subsequently applied to the large-scale industrial, experimental data to identify potentially appropriate model parameters. Results indicate that ensemble methods can accurately predict viscosity as a function of temperature and composition across all three industrial datasets. Glass density prediction shows reliable learning performance with predictions primarily being within the experimental uncertainty of the test data. Furthermore, machine learning can predict glass dissolution model parameters behavior, demonstrating potential value in GRAAL model development and in assessing suitable model parameters for large-scale industrial glass dissolution data.

Keywords: machine learning, predictive modelling, pattern recognition, radioactive waste glass

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Authors: Yeshanew Ayele Tiruneh, L. M. Modiba, S. M. Zuma

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Introduction- Healthcare waste is any waste generated by health care facilities, considered potentially hazardous to health. Solid health care waste is categorised into infectious and non-infectious wastes. Infectious waste is material suspected to contain pathogens. The non-infectious waste includes wastes that have not been in contact with infectious agents, hazardous chemicals, or radioactive substances. The purpose is to assess solid health care waste (SHCW) management practice toward developing guidelines. The setting is all health facilities found in Hossaena town. A mixed-method study design used. For the qualitative part, small purposeful samples were considered and large samples for the quantitative phase. Both samples were taken from the same population. Result - 17(3.1%) of health facility workers have hand washing facilities. 392 (72.6%) of the participants agree on the availability of one or more of personal protective equipment (PPE) in the facility ‘’the reason for the absence of some of the PPEs like boots, goggles, and shortage of disposable gloves are owing to cost inflation from time to time and sometimes absent from the market’’. The observational finding shows that colour coded waste bins are available at 23 (9.6%) of the rooms. Majority of the sharp container used in the health facility are reusable in the contrary to the health care waste management standards and most of them are plastic buckets and easily cleanable. All of the health facility infectious waste are collected transported and deposed daily. Regarding the preventive vaccination nearly half of the the fahealth facility workers wer vaccinated for Hep B virus. Conclusion- Hand washing facilities, personal protective equipment’s and preventive vaccinations are not easily available for health workers. Solid waste segregation practices are poor and these practices showed that SWMP is below the acceptable level.

Keywords: health care waste, waste management, disposal, private health facilities

Procedia PDF Downloads 74

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: 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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Authors: Peter Ekene Egwu

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Waste management is of critical importance in Africa for reasons related to public health, human dignity, climate resilience and environmental preservation. However, delivering waste management services requires adequate funding, which has generally been lacking in a context where the generation of waste is outpacing the development of waste management infrastructure in most cities. The sector represents a growing percentage of cities’ greenhouse gas (GHG) emissions, and some of the African cities profiled in this study are now designing waste management strategies with emission reduction in mind.

Keywords: management waste material, Africa, uses of new technology to manage waste, waste management

Procedia PDF Downloads 76

Authors: Zubair Ahmad

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The aim of this study is to analyze the risks factors and issues with waste management in developing nations as barriers. Depending on their content and categorization, wastes are managed differently. Waste management strategies differ for liquid, solid, and organic wastes. The final stage of trash disposal entails procedures like burning, interment, recycling, and treatment. Due to the rising creation of solid waste, the growing urban population has a magnified impact on the environment and public health. All regions, but especially informal urban neighborhoods, tribal villages, and official rural settlements have a protracted backlog in waste services. Another significant impediment seen in the developing world is a lack of education and awareness of effective waste-management practices. Unauthorized dumpsites pose a serious risk to the environment since they could contain dangerous elements like radioactive, infectious, and toxic waste. Wealthier individuals are more inclined to think that their actions will have an impact on environmental problems and to act to address them. Waste managers need to take action to make sure the public is given information that is consistent with what they currently know. The results of the data analysis conducted with the aid of the various methodologies discussed in the preceding chapter are presented in this chapter by the researcher. Descriptive analysis has been used in research to determine whether or not there are relationships between variables and to determine the importance of the variables. According to a survey, there are no efforts being made to lessen the odor that garbage dump sites emit (in terms of treating or recycling the material placed at dumpsite) This might be the case since respondents only commented on the waste management conditions in their immediate surroundings and may not have fully understood the steps taken to resolve this issue.

Keywords: risk factor of waste material, lack of awareness, developing countries struggles, waste management

Procedia PDF Downloads 71

Authors: Vladimira Vytlacilova

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Recycling of construction and demolition waste (C&DW) and their new reuse in structures is one of the solutions of environmental problems. Construction and demolition waste creates a major portion of total solid waste production in the world and most of it is used in landfills all the time. The paper deals with the situation of the recycling of the building and demolition waste in the Czech Republic during the recent years. The paper is dealing with questions of C&D waste recycling, it also characterizes construction and demolition waste in general, furthermore it analyses production of construction waste and subsequent production of recycled materials.

Keywords: Recycling, Construction and demolition waste, Recycled rubble, Waste management

Procedia PDF Downloads 303