Search results for: water cooled nuclear reactors

Authors: Taehyun Kim, Hyunjoo Park, Taehyun Kim

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Public deliberation committee on Shin-Gori Nuclear Reactors No. 5 & 6 in South Korea recently suggested policy recommendation in July 2017 including complementary measures for resumption of construction: 1) nuclear power generation reduction, 2) expansion of investment to increase proportion of renewable energy, 3) repositories of spent nuclear fuel. Even when constructing eco-friendly renewable energy facilities such as solar and wind power plants, local residents are opposed to construction of these facilities due to environmental pollution and health impacts. In order to transform eco-friendly energy, it is necessary to convert nuclear energy into renewable energy and to take measures to increase the acceptance of residents through the participation of citizens. Therefore, this study aims to compare the factors of local acceptance of renewable energy facilities and spent nuclear fuel repositories through literature review and in-depth interview. The results show that environmental and economic concerns, risk perceptions, sociality, demographic characteristics and subjective recognition types affect the local acceptance for spent nuclear fuel repository. The factors of local acceptance for renewable energy facilities are partially coincide with those for spent nuclear fuel repository. The results of this study will contribute to improving residents' acceptance and reducing conflicts when determining the location of facilities in the future.

Keywords: local acceptance, renewable energy facility, spent nuclear fuel repository, interview

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Authors: Sara Martín, Ying Jie Zheng, César Hueso

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International shipping is considered one of the largest sources of pollution in the world, accounting for 812 million tons of CO2 emissions in the year 2018. Current maritime decarbonisation is based on the implementation of new fuel alternatives, such as LNG, biofuels, and methanol, among others, which are less polluting as well as less efficient. Despite being a carbon-free and highly-developed technology, nuclear propulsion is hardly discussed as an alternative. Scientifically, it is believed that Small Modular Reactors (SMR) could be a promising solution to decarbonized maritime traffic due to their small dimensions and safety capabilities. However, as of today, there are no merchant ships powered by nuclear systems. Therefore, this project aims to understand the challenges of the development of nuclear-fuelled vessels by analysing all SMR designs to choose the most suitable one. In order not to fall into subjectivities, the Analytic Hierarchy Process (AHP) will be used to make the selection. This multiple-criteria evaluation technique analyses complex decisions by pairwise comparison of a number of evaluation criteria that can be applied to each SMR. The state-of-the-art 72 SMRs presented by the International Atomic Energy Agency (IAEA) will be analysed and ranked by a global parameter, calculated by applying the AHP methodology. The main target of the work is to find an adequate SMR system to power a ship. Top designs will be described in detail, and conclusions will be drawn from the results. This project has been conceived as an effort to foster the near-term development of zero-emission maritime traffic.

Keywords: international shipping, decarbonization, SMR, AHP, nuclear-fuelled vessels

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Authors: D. S. Gomes, A. T. Silva

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Analysis of the uncertainty quantification related to nuclear safety margins applied to the nuclear reactor is an important concept to prevent future radioactive accidents. The nuclear fuel performance code may involve the tolerance level determined by traditional deterministic models producing acceptable results at burn cycles under 62 GWd/MTU. The behavior of nuclear fuel can simulate applying a series of material properties under irradiation and physics models to calculate the safety limits. In this study, theoretical predictions of nuclear fuel failure under transient conditions investigate extended radiation cycles at 75 GWd/MTU, considering the behavior of fuel rods in light-water reactors under reactivity accident conditions. The fuel pellet can melt due to the quick increase of reactivity during a transient. Large power excursions in the reactor are the subject of interest bringing to a treatment that is known as the Fuchs-Hansen model. The point kinetic neutron equations show similar characteristics of non-linear differential equations. In this investigation, the multivariate logistic regression is employed to a probabilistic forecast of fuel failure. A comparison of computational simulation and experimental results was acceptable. The experiments carried out use the pre-irradiated fuels rods subjected to a rapid energy pulse which exhibits the same behavior during a nuclear accident. The propagation of uncertainty utilizes the Wilk's formulation. The variables chosen as essential to failure prediction were the fuel burnup, the applied peak power, the pulse width, the oxidation layer thickness, and the cladding type.

Keywords: logistic regression, reactivity-initiated accident, safety margins, uncertainty propagation

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Authors: Massimo Zucchetti

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In the US, the SPARC-ARC projects of compact tokamaks are being developed: both are aimed at the technological demonstration of fusion power reactors with cutting-edge technology but following different design approaches. However, they show more similarities than differences in the fuel cycle, safety, radiation protection, environmental, waste and decommissioning aspects: all reactors, either experimental or demonstration ones, have to fulfill certain "essential" requirements to pass from virtual to real machines, to be built in the real world. The paper will discuss these "essential" requirements. Some of the relevant activities in these fields, carried out by our research group (ESSENTIAL group), will be briefly reported, with the aim of showing some methodology aspects that have been developed and might be of wider interest. Also, a non-competitive comparison between our results for different projects will be included when useful. The question of advanced D-He3 fuel cycles to be used for those machines will be addressed briefly. In the past, the IGNITOR project of a compact high-magnetic field D-T ignition experiment was found to be able to sustain limited D-He3 plasmas, while the Candor project was a more decisive step toward D-He3 fusion reactors. The following topics will be treated: Waste management and radioactive safety studies for advanced fusion power plants; development of compact high-field advanced fusion reactors; behavior of nuclear materials under irradiation: neutron-induced radioactivity due to side DT reactions, radiation damage; accident analysis; reactor siting.

Keywords: advanced fuel fusion reactors, deuterium-helium3, high-field tokamaks, fusion safety

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Authors: A. Roukbi, B. Draoui

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Due to the high cost and the environmental problems caused by the conventional air-conditioning systems, various researches are being increasingly focused on thermal comfort in the building sector integrating renewable energy sources, particularly solar energy. For that purpose, this study aims to present a modeling and performance analysis of a direct air-cooled Water/LiBr absorption chiller. The chiller is considered to be coupled to a small residential building at an arid zone situated in south Algeria. The system is modeled with TRNSYS simulation program. The main objective is to study the feasibility of the chosen system in arid zones and to apply a simplified method to predict the performance of the system by mean of the characteristic equation approach tacking in account the influence of the climatic conditions of the considered site, the collector area and storage volume of the hot water tank on the performance of the installation. First, the results of the system modeling are compared with an experimental data from the open literature and the developed model is then validated. In another hand, a parametric study is performed to analyze the performance of the direct air-cooled absorption chiller at the operating conditions of interest for the present study. Thus, the obtained results has shown that the studied system can present a good alternative for cooling systems in arid zones since the cooling load is roughly in phase with solar availability.

Keywords: absorption chiller, air-cooled, arid zone, thermal comfort

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Authors: H. Shokouhmand, H. Sobhani, B. Sajadi, M. Degheh

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In recent years, experimental and numerical investigation of water UV reactors has increased significantly. The main drawback of experimental methods is confined and expensive survey of UV reactors features. In this study, a CFD model utilizing the eulerian-lagrangian framework is applied to analysis the disinfection performance of a closed conduit reactor which contains four UV lamps perpendicular to the flow. A discrete ordinates (DO) model was employed to evaluate the UV irradiance field. To investigate the importance of each of lamps on the inactivation performance, in addition to the reference model (with 4 bright lamps), several models with one or two bright lamps in various arrangements were considered. All results were reported in three inactivation kinetics. The results showed that the log inactivation of the two central bright lamps model was between 88-99 percent, close to the reference model results. Also, whatever the lamps are closer to the main flow region, they have more effect on microbial inactivation. The effect of some operational parameters such as water flow rate, inlet water temperature, and lamps power were also studied.

Keywords: Eulerian-Lagrangian framework, inactivation kinetics, log inactivation, water UV reactor

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Authors: E. A. Krasikov

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As the service life of an operating nuclear power plant (NPP) increases, the potential misunderstanding of the degradation of aging components must receive more attention. Integrity assurance analysis contributes to the effective maintenance of adequate plant safety margins. In essence, the reactor pressure vessel (RPV) is the key structural component determining the NPP lifetime. Environmentally induced cracking in the stainless steel corrosion-preventing cladding of RPV’s has been recognized to be one of the technical problems in the maintenance and development of light-water reactors. Extensive cracking leading to failure of the cladding was found after 13000 net hours of operation in JPDR (Japan Power Demonstration Reactor). Some of the cracks have reached the base metal and further penetrated into the RPV in the form of localized corrosion. Failures of reactor internal components in both boiling water reactors and pressurized water reactors have increased after the accumulation of relatively high neutron fluences (5´1020 cm–2, E>0,5MeV). Therefore, in the case of cladding failure, the problem arises of hydrogen (as a corrosion product) embrittlement of irradiated RPV steel because of exposure to the coolant. At present when notable progress in plasma physics has been obtained practical energy utilization from fusion reactors (FR) is determined by the state of material science problems. The last includes not only the routine problems of nuclear engineering but also a number of entirely new problems connected with extreme conditions of materials operation – irradiation environment, hydrogenation, thermocycling, etc. Limiting data suggest that the combined effect of these factors is more severe than any one of them alone. To clarify the possible influence of the in-service synergistic phenomena on the FR structural materials properties we have studied hydrogen-irradiated steel interaction including alternating hydrogenation and heat treatment (annealing). Available information indicates that the life of the first wall could be expanded by means of periodic in-place annealing. The effects of neutron fluence and irradiation temperature on steel/hydrogen interactions (adsorption, desorption, diffusion, mechanical properties at different loading velocities, post-irradiation annealing) were studied. Experiments clearly reveal that the higher the neutron fluence and the lower the irradiation temperature, the more hydrogen-radiation defects occur, with corresponding effects on the steel mechanical properties. Hydrogen accumulation analyses and thermal desorption investigations were performed to prove the evidence of hydrogen trapping at irradiation defects. Extremely high susceptibility to hydrogen embrittlement was observed with specimens which had been irradiated at relatively low temperature. However, the susceptibility decreases with increasing irradiation temperature. To evaluate methods for the RPV’s residual lifetime evaluation and prediction, more work should be done on the irradiated metal–hydrogen interaction in order to monitor more reliably the status of irradiated materials.

Keywords: hydrogen, radiation, stability, structural steel

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Authors: Kenichi Hirota, Jifeng Wang, Sadao Araki, Koji Endo, Hideki Yamamoto

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Zirconium (Zr) has been used as a fuel cladding tube for nuclear reactors, because of the excellent corrosion resistance and the low adsorptive material for neutron. Generally speaking, the natural resource of Zr is often containing Hf that has similar properties. The content of Hf in the Zr resources is about 2~4 wt%. In the industrial use, the content of Hf in Zr resources should be lower than the 100 ppm. However, the separation of Zr and Hf is not so easy, because of similar chemical and physical properties such as melting point, boiling point and things. Solvent extraction method has been applied for the separation of Zr and Hf from Zr natural resources. This method can separate Hf with high efficiency (Hf < 100ppm), however, it needs much amount of organic solvents for solvent extraction and the cost of its disposal treatment is high. Therefore, we attached attention for the fractional crystallization. This separation method depends on the solubility difference of Zr and Hf in the solvent. In this work, hexafluorozirconate (hafnate) (K2Zr(Hf)F6) was used as model compound. Solubility of K2ZrF6 in water showed lower than that of K2HfF6. By repeating of this treatment, it is possible to purify Zr, practically. In this case, 16-18 times of recrystallization stages were needed for its high purification. The improvement of the crystallization process was carried out in this work. Water, hydrofluoric acid (HF) and hydrofluoric acid (HF) +hydrochloric acid (HCl) mixture were chosen as solvent for dissolution of Zr and Hf. In the experiment, 10g of K2ZrF6 was added to each solvent of 100mL. Each solution was heated for 1 hour at 353K. After 1h of this operation, they were cooled down till 293K, and were held for 5 hours at 273K. Concentration of Zr or Hf was measured using ICP analysis. It was found that Hf was separated from Zr-Hf mixed compound with high efficiency, when HF-HCl solution was used for solvent of crystallization. From the comparison of the particle size of each crystal by SEM, it was confirmed that the particle diameter of the crystal showed smaller size with decreasing of Hf content. This paper concerned with purification of Zr from Zr-Hf mixture using crystallization method.

Keywords: crystallization, zirconium, hafnium, separation

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Authors: Saeed Vahidifar, Mohammad Nakhaee Sharif, Mohammad Ghaffari

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Few people would dispute the fact that one of the most common applications of energy is creating comfort in buildings, so it is probably true to say that management of energy consumption is required due to the environmental issues and increasing the efficiency of mechanical systems. From the geographical point of view, Iran is located in a warm and semi-arid region; therefore, air-cooled chillers are usually used for cooling residential buildings, commercial buildings, medical buildings, etc. In this study, a heat exchanger was designed for providing laundry hot water by utilizing condenser heat lost base on analytical results of a 540-bed hospital in the city of Mashhad in Iran. In this paper, by using the analytical method, energy consumption reduces about 13%, and coefficient of performance increases a bit. Results show that this method can help in the management of energy consumption a lot.

Keywords: air cooled chiller, energy management, environmental issues, heat exchanger, hospital laundry system

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Authors: Liang Zhang, Gang Xu, Yue Wang, Chen Li, Shao Chong Zhou

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Presently, more and more nuclear power plants are facing the issue of life extension. When a nuclear power plant applies for an extension of life, its condition needs to meet the current design standards, which is not fine for all old reactors, typically for seismic design. Seismic-grade equipment in nuclear power plants are now generally placed separately from the non-seismic-grade equipment, but it was not strictly required before. Therefore, it is very important to study whether non-seismic-grade equipment will affect the seismic-grade equipment when dropped down in an earthquake situation, which is related to the safety of nuclear power plants and future life extension applications. This research was based on the cooling water system with the seismic and non-seismic grade equipment installed together, as an example to study whether the non-seismic-grade equipment such as DN50 fire pipes and ventilation pipes arranged above will damage the DN300 pipes and valves arranged below when earthquakes occur. In the study, the simulation was carried out by ANSYS / LY-DYNA, and Johnson-Cook was used as the material model and failure model. For the experiments, the relative positions of objects in the room were restored by 1: 1. In the experiment, the pipes and valves were filled with water with a pressure of 0.785 MPa. The pressure-holding performance of the pipe was used as a criterion for damage. In addition to the pressure-holding performance, the opening torque was considered as well for the valves. The research results show that when the 10-meter-long DN50 pipe was dropped from the position of 8 meters height and the 8-meter-long air pipe dropped from a position of 3.6 meters height, they do not affect the integrity of DN300 pipe below. There is no failure phenomenon in the simulation as well. After the experiment, the pressure drop in two hours for the pipe is less than 0.1%. The main body of the valve does not fail either. The opening torque change after the experiment is less than 0.5%, but the handwheel of the valve may break, which affects the opening actions. In summary, impacts of the upper pipes and ventilation pipes dropdown on the integrity of the DN300 pipes and valves below in a cooling water system of a typical second-generation nuclear power plant under an earthquake was studied. As a result, the functionality of the DN300 pipeline and the valves themselves are not significantly affected, but the handwheel of the valve or similar articles can probably be broken and need to take care.

Keywords: cooling water system, earthquake, integrity, pipe and valve

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Authors: Syazwani Idrus, Charles Banks, Sonia Heaven

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The effect of osmotic stress was carried out to determine the ability for biogas production in two types of digesters; anaerobic sludge blanket and anaerobic filters in treating wheat straw washed water. Two anaerobic filters (AF1 and 2) and two UASB reactors (U1 and 2) with working volumes of 1.5 L were employed at mesophilic temperatures (37°C). Digesters AF1 and two were seeded with an inoculum which had previously been fed on with a synthetic wastewater includingSodium Chloride and Potassium Chloride. Digesters U1 and two were seeded with 1 kg wet weight of granular sludge which had previously been treating paper mill effluent. During the first 48 days, all digesters were successfully acclimated with synthetic wastewater (SW) to organic loading rate (OLR) of 6 g COD l^-1 day-1. Specific methane production (SMP) of 0.333 l CH4 g-1 COD). The feed was then changed to wash water from a washing operation to reduce the salt content of wheat straw (wheat straw wash water, WSW) at the same OLR. SMP fell sharply in all reactors to less than 0.1 l CH4 g^-1 COD, with the AF affected more than the UASB. The OLR was reduced to 2.5 g COD l^-1 day^-1 to allow adaptation to WSW, and both the UASB and the AF reactors achieved an SMP of 0.21 l CH4 g^-1 COD added at 82% of COD removal. This study also revealed the accumulation of potassium (K) inside the UASB granules to a concentration of 4.5 mg K g^-1 wet weight of granular sludge. The phenomenon of lower SMP and accumulation of K indicates the effect of osmotic stress when fed on WSW. This finding is consistent with the theory that methanogenic organisms operate a Potassium pump to maintain ionic equilibrium, and as this is an energy-driven process, it will, therefore, reduce the overall methane yield.

Keywords: wheat straw wash water, upflow anaerobic sludge blanket, anaerobic filter, specific methane production, osmotic stress

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Authors: Pinar Temocin

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Although the Akkuyu nuclear power project has been in the planning the Akkuyu nuclear power plant in the Mersin Province of Southern Turkey, recent events have increased its visibility in the Turkish debate. The Fukushima accident, the 2010 nuclear deal with Russia followed by several consequent nuclear revelations of administrative deficiencies, and waste issues all spurted widespread protests across Turkey and have polarized the nation into two camps; supporters and detractors. Those who support a nuclear Turkey include energy entrepreneurs, local investors, and technical experts who are heavily involved in paving the way for the realization of a nuclear project. Civil society activists and environmentalists overwhelmingly oppose the nuclear program. This study focuses on the latter, analyzing how groups opposing nuclear power plants (NPPs) have framed the Akkuyu nuclear project as a dangerous, risky, disadvantageous, and irrational policy choice.

Keywords: nuclear energy, anti-nuclear movements, environmentalists, civil society, Turkey

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Authors: Jong-Sung Kim, Myung-Jo Jhung

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BFA is one of the reactor internals components in PWR. The BFA has the intended functions to support fuel assembly, to keep structural integrity of upper/lower core support structures, and to secure reactor coolant flow path. Failure of the BFA may give rise to significant effect on reactor safety operation and stop. The BFA is subject to relatively high neutron irradiation dose due to location close to the core. Therefore, IASCC can occur on the BFA due to damage accumulation as operating year increases. In this study, IASCC susceptibility on the BFA was assessed via the FEA considering variations of mechanical material behaviors with neutron irradiation. As a result of the assessment, some points have susceptibility more than 0.2 to IASCC during design lifetime.

Keywords: baffle former assembly, finite element analysis, irradiation aging, nuclear power plant, pressurized water reactor

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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: Irfan Turetgen

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Closed-circuit cooling towers are cooling units that operate according to the indirect cooling principle. Unlike the open-loop cooling tower, the filler material includes a closed-loop water-operated heat exchanger. The main purpose of this heat exchanger is to prevent the cooled process water from contacting with the external environment. In order to ensure that the hot water is cooled, the water is cooled by the air flow and the circulation water of the tower as it passes through the pipe. They are now more commonly used than open loop cooling towers that provide cooling with plastic filling material. As with all surfaces in contact with water, there is a biofilm formation on the outer surface of the pipe. Although biofilm has been studied very well on plastic surfaces in open loop cooling towers, studies on biofilm layer formed on the heat exchangers of the closed circuit tower have not been found. In the recent study, natural biofilm formation was observed on the heat exchangers of the closed loop tower for 6 months. At the same time, nano-silica coating, which is known to reduce the formation of the biofilm layer, a comparison was made between the two different surfaces in terms of biofilm formation potential. Test surfaces were placed into biofilm reactor along with the untreated control coupons up to 6-months period for biofilm maturation. Natural bacterial communities were monitored to analyze the impact to mimic the real-life conditions. Surfaces were monthly analyzed in situ for their microbial load using epifluorescence microscopy. Wettability is known to play a key role in biofilm formation on surfaces, because characteristics of surface properties affect the bacterial adhesion. Results showed that surface-conditioning with nano-silica significantly reduce (up to 90%) biofilm formation. Easy coating process is a facile and low-cost method to prepare hydrophobic surface without any kinds of expensive compounds or methods.

Keywords: biofilms, cooling towers, fill material, nano silica

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Authors: Nazanin Gerami, Shirin Adlparvar

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In the present time in the laboratory, one can create an ionized gas, that is to say, plasma from room temperature up to ten times more than the temperature of the sun center (150,000,000). All these temperature spectrums of plasma have applications in different disciplines, including dentistry, medicine, science, surface treatment, nuclear waste disinfection, nuclear fusion technology, etc. However, for the sick of simplicity, all these plasma temperature spectrums are classified as cold or low-pressure non-thermal plasma and warm or high-pressure equilibrium plasma. The cold plasma, as we are interested in this paper, exists at lower ion and neutral temperatures with respect to electron temperature, but in the equilibrium plasma, the temperatures of ion and electron are fairly equal. The cold plasma is a partially ionized gas comprising ions, electrons, ultraviolet photons and reactive neutrals such as radicals, excited and ground-state molecules. Cold atmospheric pressure plasmas are widely used in diverse fields of dental medicine, such as the titanium surface of dental implants, which helps in reducing contact angle and supporting the spread of osteoblastic cells and is known to aid in osteoblastic proliferation and osseointegration, thus increasing the success rates of implants. This article focuses on the anticipated uses of a newly designed water-cooled adjustable cathode cold atmospheric pressure plasma Jet (CAPPJ) for titanium surface treatment in dental implant placement.

Keywords: CAPPJ, surface modification, osseointegration, plasma medicine, dentistry

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Authors: Jana Petru, Marie Kudrnova

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The use of technologies working with molten salts is conditioned by finding suitable construction materials that must meet several demanding criteria. In addition to temperature resistance, materials must also show corrosion resistance to salts; they must meet mechanical requirements and other requirements according to the area of use – for example, radiation resistance in Molten Salt Reactors. The present text describes an experimental device for studying the corrosion resistance of candidate materials in molten mixtures of salts and is a partial task of the international project ADAR, dealing with the evaluation of advanced nuclear reactors based on molten salts. The design of the device is based on a test exposure of Inconel 625 in the mixture of salts Hitec in a high temperature tube furnace. The result of the pre-exposure is, in addition to the metallographic evaluation of the behavior of material 625 in the mixture of nitrate salts, mainly a list of operational and construction problems that were essential for the construction of the new experimental equipment. The main output is a scheme of a newly designed gas-tight experimental apparatus capable of operating in an inert argon atmosphere, temperature up to 600 °C, pressure 3 bar, in the presence of a corrosive salt environment, with an exposure time of hundreds of hours. This device will enable the study of promising construction materials for nuclear energy.

Keywords: corrosion, experimental device, molten salt, steel

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Authors: Caroline E. Mendes, Alberto C. Badino

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Pneumatic reactors have been widely employed in various sectors of the chemical industry, especially where are required high heat and mass transfer rates. This study aimed to obtain correlations that allow the prediction of gas hold-up (Ԑ) and volumetric oxygen transfer coefficient (kLa), and compare these values, for three models of pneumatic reactors on two scales utilizing Newtonian fluids. Values of kLa were obtained using the dynamic pressure-step method, while  was used for a new proposed measure. Comparing the three models of reactors studied, it was observed that the mass transfer was superior to draft-tube airlift, reaching  of 0.173 and kLa of 0.00904s-1. All correlations showed good fit to the experimental data (R2≥94%), and comparisons with correlations from the literature demonstrate the need for further similar studies due to shortage of data available, mainly for airlift reactors and high viscosity fluids.

Keywords: bubble column, internal loop airlift, gas hold-up, kLa

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Authors: Smita Singh

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Through the constructivist perspective, this paper explores the transformation of India’s nuclear identity from an irresponsible nuclear weapon power to a ‘de-facto nuclear power’ in the emerging international nuclear order From a nuclear abstainer to a bystander and finally as a ‘de facto nuclear weapon state’, India has put forth its case as a unique and exceptional nuclear power as opposed to Iran, Iraq and North Korea with similar nuclear ambitions, who have been snubbed as ‘rogue states’ by the international community. This paper investigates the reasons behind international community’s gradual acceptance of India’s nuclear weapons capabilities and nuclear identity after the Indo-U.S. Nuclear Deal. In this paper, the central concept of analysis is the inter-subjective nature of identity in the nuclear arena. India’s nuclear behaviour has been discursively constituted by India through evolving images of the ‘self’ and the ‘other.’ India’s sudden heightened global status is not solely the consequence of its 1998 nuclear tests but a calibrated projection as a responsible stakeholder in other spheres such as economic potential, market prospects, democratic credentials and so on. By examining India’s nuclear discourse this paper contends that India has used its material and discursive power in presenting a n striking image as a responsible nuclear weapon power (though not yet a legal nuclear weapon state as per the NPT). By historicising India’s nuclear trajectory through an inter-subjective analysis of identities, this paper moves a step ahead in providing a theoretical interpretation of state actions and nuclear identity construction.

Keywords: nuclear identity, India, constructivism, international stakeholder

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Authors: Bharat Kumar, Deepak Kumar, Vijay Chaudhry

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Zirconium alloys are used as core components of nuclear reactors due to their high wear resistance, good corrosion properties, and good mechanical stability at high temperatures. Water flows inside the pressure tube through fuel claddings, which produces vibration of these core components and results in the wear of some components. Some components are subjected to the environment of coolant water containing LiOH which results in the corrosion of these components. The present work simulates some of these conditions to determine the failure mechanisms under these conditions and the effect of various parameters on them. Friction and wear experiments were performed varying the surrounding environment (room temperature, high temperature, and water submerged), duration, frequency, and displacement amplitude. Electrochemical corrosion experiments were performed by varying the concentration of LiOH in water. The worn and corroded surfaces were analyzed using scanning electron microscopy (SEM) to analyze the wear and corrosion mechanism and energy dispersive x-ray spectroscopy (EDS) and Raman spectroscopy to analyze the tribo-oxide layer formed during the wear and oxide layer formed during the corrosion. Wear increases with frequency and amplitude, and corrosion increases with LiOH concentration in water.

Keywords: zirconium alloys, wear, oxide layer, corrosion, EIS, linear polarization

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Authors: Eitidal Albassam

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Cooling towers (CTs) in arid zone countries, used for heat rejection in water-cooled (WC) systems, consume a large quantity of water. Universally, water conservation is an issue because of the scarcity of fresh water and natural resources. Therefore, many studies have aimed to conserve fresh water and limit the water wasted. Nonetheless, all these methods are not related to improving the weather conditions around the entering air to CT. In Kuwait and other arid-zone countries, the dry bulb temperature (DBT) during the summer season is significantly greater than the incoming hot water temperature, and the air undergoes sensible cooling. This high DBT leads to extra heat transfer from air to water, demanding high water vaporization to achieve the required cooling of water. Thus, any reduction in ambient air temperature around the CT will minimize water consumption. This paper aims to discuss theoretically the effect of reducing the DBT around the cooling tower when considering the sun-shading barrier. The theoretical simulation model results show that if the DBT reduces by 2.8 °C approximately, the percentage of water evaporation savings in gallon per minute (GPM) starts from 6.48% when DBT reaches 51.67 °C till 9.80% for 37.78 °C. Moreover, the performance of the cooling tower will be improved when considering the appropriate shading barriers, which will not affect the existing wet-bulb temperature.

Keywords: dry-bulb temperature, entering air, water consumption, water vaporization

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Authors: Radek Novotny, Michal Novak, Daniela Marusakova, Monika Sipova, Hugo Fuentes, Peter Borst

Abstract:

This research has been conducted within the European HORIZON 2020 project ECC-SMART. The main objective is to assess whether it is feasible to design and develop a small modular reactor (SMR) that would be cooled by supercritical water (SCW). One of the main objectives for material research concerns the corrosion of the candidate cladding materials. The experimental part has been conducted in support of the qualification procedure of the future SCW-SMR constructional materials. The last objective was to identify the gaps in current norms and guidelines. Apart from corrosion, resistance testing of candidate materials stresses corrosion cracking susceptibility tests have been performed in supercritical water. This paper describes part of these tests, in particular, those slow strain rate tensile loading applied for tangential ring shape specimens of two candidate materials, Alloy 800H and 310S stainless steel. These ring tensile tests are one the methods used for tensile testing of nuclear cladding. Here full circular heads with dimensions roughly equal to the inner diameter of the sample and the gage sections are placed in the parallel direction to the applied load. Slow strain rate tensile tests have been conducted in 380 or 500oC supercritical water applying two different elongation rates, 1x10-6 and 1x10-7 s-1. The effect of temperature and dissolved oxygen content on the SCC susceptibility of Alloy 800H and 310S stainless steel was investigated when two different temperatures and concentrations of dissolved oxygen were applied in supercritical water. The post-fracture analysis includes fractographic analysis of the fracture surfaces using SEM as well as cross-sectional analysis on the occurrence of secondary cracks. Assessment of the effect of environment and dissolved oxygen content was by comparing to the results of the reference tests performed in air and N2 gas overpressure. The effect of high temperature on creep and its role in the initiation of SCC was assessed as well. It has been concluded that the applied test method could be very useful for the investigation of stress corrosion cracking susceptibility of candidate cladding materials in supercritical water.

Keywords: stress corrosion cracking, ring tensile tests, super-critical water, alloy 800H, 310S stainless steel

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Authors: Sam Rasoulzadeh, Atefeh Mousavi

Abstract:

Fluidized bed reactors are known as highly exothermic and endothermic according to uniformity in temperature as a safe and effective mean for catalytic reactors. In these reactors, a wide range of catalyst particles can be used and by using a continuous operation proceed to produce in succession. Providing optimal conditions for the operation of these types of reactors will prevent the exorbitant costs necessary to carry out laboratory work. In this regard, a hydrodynamic analysis was carried out with heat transfer in the solid-gas fluidized bed reactor for solar thermal applications. The results showed that in the fluid flow the input of the reactor has a lower temperature than the outlet, and when the fluid is passing from the reactor, the heat transfer happens between cylinder and solar panel and fluid. It increases the fluid temperature in the outlet pump and also the kinetic energy of the fluid has been raised in the outlet areas.

Keywords: heat transfer, solar reactor, fluidized bed reactor, CFD, computational fluid dynamics

Procedia PDF Downloads 178

Authors: Imane Khalil, Quinn Pratt

Abstract:

In a nuclear reactor, an array of fuel rods containing stacked uranium dioxide pellets clad with zircalloy is the heat source for a thermodynamic cycle of energy conversion from heat to electricity. After fuel is used in a nuclear reactor, the assemblies are stored underwater in a spent nuclear fuel pool at the nuclear power plant while heat generation and radioactive decay rates decrease before it is placed in packages for dry storage or transportation. A computational model of a Boiling Water Reactor spent fuel assembly is modeled using FLUENT, the computational fluid dynamics package. Heat transfer simulations were performed on the two-dimensional 9x9 spent fuel assembly to predict the maximum cladding temperature for different input to the FLUENT model. Uncertainty quantification is used to predict the heat transfer and the maximum temperature profile inside the assembly.

Keywords: spent nuclear fuel, conduction, heat transfer, uncertainty quantification

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Authors: Harish Aryal

Abstract:

The safe and reliable operation of nuclear reactors has always been one of the topmost priorities in the nuclear industry. Transient testing allows us to understand the time-dependent behavior of the neutron population in response to either a planned change in the reactor conditions or unplanned circumstances. These unforeseen conditions might occur due to sudden reactivity insertions, feedback, power excursions, instabilities, and accidents. To study such behavior, we need transient testing, which is like car crash testing, to estimate the durability and strength of a car design. In nuclear designs, such transient testing can simulate a wide range of accidents due to sudden reactivity insertions and helps to study the feasibility and integrity of the fuel to be used in certain reactor types. This testing involves a high neutron flux environment and real-time imaging technology with advanced instrumentation with appropriate accuracy and resolution to study the fuel slumping behavior. With the aid of transient testing and adequate imaging tools, it is possible to test the safety basis for reactor and fuel designs that serves as a gateway in licensing advanced reactors in the future. To that end, it is crucial to fully understand advanced imaging techniques both analytically and via simulations. This paper presents an innovative method of supporting real-time imaging of fuel pins and other structures during transient testing. The major fuel-motion detection device that is studied in this dissertation is the Hodoscope which requires collimators. This paper provides 1) an MCNP model and simulation of a Transient Reactor Test (TREAT) core with a central fuel element replaced by a slotted fuel element that provides an open path between test samples and a hodoscope detector and 2) a choice of good filter to improve image resolution.

Keywords: hodoscope, transient testing, collimators, MCNP, TREAT, hodogram, filters

Procedia PDF Downloads 77

Authors: Assabo Mohamed, Bile Mohamed, Ali Farah, Isman Souleiman, Olga Alos Ramos, Marie Cadet

Abstract:

The life of a nuclear power plant is regularly punctuated by short or long period outages to carry out maintenance operations and/or nuclear fuel reloading. During these stops periods, it is essential to conserve all the secondary circuit equipment to avoid corrosion priming. This kind of circuit is one of the main components of a nuclear reactor. Indeed, the conservation materials on shutdown of a nuclear unit improve circuit performance and reduce the maintenance cost considerably. This study is a part of the optimization of the dry preservation of equipment from the water station of the nuclear reactor. The main objective is to provide tools to guide Electricity Production Nuclear Centre (EPNC) in order to achieve the criteria required by the chemical specifications of conservation materials. A theoretical model of drying exchangers of water station is developed by the software Engineering Equation Solver (EES). It used to size requirements and air quality needed for dry conservation of equipment. This model is based on heat transfer and mass transfer governing the drying operation. A parametric study is conducted to know the influence of aerothermal factor taking part in the drying operation. The results show that the success of dry conservation of equipment of the secondary circuit of nuclear reactor depends strongly on the draining, the quality of drying air and the flow of air injecting in the secondary circuit. Finally, theoretical case study performed on EES highlights the importance of mastering the entire system to balance the air system to provide each exchanger optimum flow depending on its characteristics. From these results, recommendations to nuclear power plants can be formulated to optimize drying practices and achieve good performance in the conservation of material from the water at the stop position.

Keywords: dry conservation, optimization, sizing, water station

Procedia PDF Downloads 262

Authors: Ali Ghobadi, Eisa Bakhoda, Hamid R. Javdan

Abstract:

One of the key factors in air cooled heat exchangers operation is the proper control of process stream outlet temperature. In this study, performances of two different air cooled heat exchangers have been considered, one of them condenses Propane and the other one cools LPG streams. In order to predict operation of these air coolers at different operating conditions. The results of simulations were applied for both economical evaluations and operational considerations for using convenient air cooler control system. In this paper, using On-Off fans method and installing variable speed drivers have been studied. Finally, the appropriate methods for controlling outlet temperature of process fluid streams as well as saving energy consumption were proposed. Using On-Off method for controlling studied Propane condenser by multiple fans is proper; while controlling LPG air cooler with lesser fans by means of two variable speed drivers is economically convenient.

Keywords: air cooled heat exchanger, simulation, economical evaluation, energy, process control

Procedia PDF Downloads 411

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

Abstract:

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

Procedia PDF Downloads 268

Authors: Kihyon Kwon, Hyun T. Park, Gil Y. Chung, Sang-Hoon Lee

Abstract:

This paper focuses on assessing sloshing-induced overflow of the seismically-isolated nuclear tanks based on Fluid-Structure Interaction (FSI) analysis. Typically, fluid motion in the seismically-isolated nuclear tank systems may be rather amplified and even overflowed under earthquake. Sloshing-induced overflow in those structures has to be reliably assessed and predicted since it can often cause critical damages to humans and environments. FSI analysis is herein performed to compute the total cumulative overflowed water volume more accurately, by coupling ANSYS with CFX for structural and fluid analyses, respectively. The approach is illustrated on a nuclear liquid storage tank, Spent Fuel Pool (SFP), forgiven conditions under consideration: different liquid levels, Peak Ground Accelerations (PGAs), and post earthquakes.

Keywords: FSI analysis, seismically-isolated nuclear tank system, sloshing-induced overflow

Procedia PDF Downloads 474

Authors: Fares Almomani, Majeda Khraisheh, Rahul Bhosale, Anand Kumar, Ujjal Gosh

Abstract:

Septic tanks (STs) are very commonly used wastewater collection systems in the world especially in rural areas. In this study, the use of moving bed biological reactors (MBBR) for the treatment of septic tanks effluents (STE) was studied. The study was included treating septic tank effluent from one house hold using MBBRs. Significant ammonia removal rate was observed in all the reactors throughout the 180 days of operation suggesting that the MBBRs were successful in reducing the concentration of ammonia from septic tank effluent. The average ammonia removal rate at 25◦C for the reactor operated at hydraulic retention time of 5.7 hr (R1) was 0.540 kg-N/m3and for the reactor operated at hydraulic retention time of 13.3hr (R2) was 0.279 kg-N/m3. Ammonia removal rates were decreased to 0.3208 kg-N/m3 for R1 and 0.212 kg-N/m3 for R3 as the temperature of reactor was decreased to 8 ◦C. A strong correlation exists between theta model and the rates of ammonia removal for the reactors operated in continuous flow. The average ϴ values for the continuous flow reactors during the temperature change from 8°C to 20 °C were found to be 1.053±0.051. MBBR technology can be successfully used as a polishing treatment for septic tank effluent.

Keywords: septic tanks, wastewater treatment, morphology, moving biological reactors, nitrification

Procedia PDF Downloads 342