Search results for: nuclear power reactors

Authors: Beril Tuğrul

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Electrical energy is the most important form of energy and electrical power plants have highest impact factor in energy policy. This study is in relation with evaluation of various power plants including fossil fuels, nuclear and renewable energy based power plants. The power plants evaluated with regard to their overall impact that considered for establishing of the plants. Both positive and negative impacts of power plant operation are compared view of different arguments. Then calculate the impact factor by using variation linear extrapolation for each argument. With this study, power plants assessed with the different point of view and clarified objectively.

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Procedia PDF Downloads 524

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

Procedia PDF Downloads 269

Authors: D. D. Nguyen, H. S. Park, S. W. Yang, B. Thusa, Y. M. Kim, T. H. Lee

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Currently, the design response spectrum (i.e., Nuclear Regulatory Commission - NRC 1.60 spectrum) with the peak ground acceleration (PGA) 0.3g (for Safe Shutdown Earthquake level) is specified for designing the new nuclear power plant (NPP) structures in Korea. However, the recent earthquakes in the region such as the 2016 Gyeongju and the 2017 Pohang earthquake showed that the possible PGA of ground motions can be larger than 0.3g. Therefore, there is a need to analyze the seismic performance of the existing NPP structures under these earthquakes. An NPP model, APR-1400, which is designed and built in Korea was selected for a case study. The NPP structure is numerically modeled in terms of lumped-mass stick elements using OpenSees framework. The floor acceleration and displacement of components are measured to quantify the responses of components. The numerical results show that the floor spectral accelerations are significantly amplified in the components subjected to Korean earthquakes. A comparison between floor response spectra of Korean earthquakes and the NRC design motion highlights that the seismic design level of NPP components under an earthquake should be thoroughly reconsidered. Additionally, a seismic safety assessment of the equipment and relays attached to main structures is also required.

Keywords: nuclear power plant, floor response spectra, Korean earthquake, NRC spectrum

Procedia PDF Downloads 158

Authors: Sa-Kil Kim, Jang-Soo Lee

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On March 31, 1993, a severe fire accident took place in a nuclear power plant located in Narora in North India. The event involved a major fire in the turbine building of NAPS unit-1 and resulted in a total loss of power to the unit for 17 hours. In addition, there was a heavy ingress of smoke in the control room, mainly through the intake of the ventilation system, forcing the operators to vacate the control room. The Narora fire accident provides us lessons indicating that operators could lose their mind and predictable behaviors during a fire. After the Fukushima accident, which resulted from a natural disaster, unanticipated external events are also required to be prepared and controlled for the ultimate safety of nuclear power plants. From last year, our research team has developed a test and evaluation facility that can simulate external events such as an earthquake and fire based on the operators’ real-sense. As one of the results of the project, we proposed a unit real-sense-based facility that can simulate fire events in a control room for utilizing a test-bed of human factor validation. The test-bed has the operator’s workstation shape and functions to simulate fire conditions such as smoke, heat, and auditory alarms in accordance with the prepared fire scenarios. Furthermore, the test-bed can be used for the operators’ training and experience.

Keywords: human behavior in fire, human factors validation, nuclear power plants, real-sense-based fire simulation

Procedia PDF Downloads 283

Authors: Andressa dos Santos Nicolau, João P. da S.C Algusto, Claudio Márcio do N. A. Pereira, Roberto Schirru

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In case of abnormal situations, the nuclear power plant (NPP) operators must follow written procedures to check the condition of the plant and to classify the type of emergency. In this paper, we proposed a Real Time Expert System in order to improve operator’s performance in case of transient or accident with reactor shutdown. The expert system’s knowledge is based on the sequence of events (SoE) of known accident and two emergency procedures of the Brazilian Pressurized Water Reactor (PWR) NPP and uses two kinds of knowledge representation: rule and logic trees. The results show that the system was able to classify the response of the automatic protection systems, as well as to evaluate the conditions of the plant, diagnosing the type of occurrence, recovery procedure to be followed, indicating the shutdown root cause, and classifying the emergency level.

Keywords: emergence procedure, expert system, operator support, PWR nuclear power plant

Procedia PDF Downloads 333

Authors: W. S. Hsu, Y. Chiang, H. C. Chen, J. R. Wang, S. W. Chen, J. H. Yang, C. Shih

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Not only radiation materials, but also the normal chemical material stored in the power plant can cause a risk to the residents. In this research, the ALOHA code was used to perform the concentration analysis under the CO₂ storage burst or leakage conditions for Kuosheng nuclear power plant (NPP). The Final Safety Analysis Report (FSAR) and data were used in this study. Additionally, the analysis results of ALOHA code were compared with the R.G. 1.78 failure criteria in order to confirm the control room habitability. The comparison results show that the ALOHA result for burst case was 0.923 g/m³ which was below the criteria. However, the ALOHA results for leakage case was 11.3 g/m³.

Keywords: BWR, ALOHA, habitability, Kuosheng

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Authors: Aiswarya C. L, Swatantra Pratap Singh

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In recent years, most desalination plants have been powered by fossil fuels, and to a lesser extent, by green energy. Greenhouse gases emitted by fossil-fuelled plants significantly impact the global climate. So scientists are forced to develop a powerful energy source to protect the environment with greater sustainability due to climate change issues. Nuclear energy can supply much more fresh water than what is currently available. Furthermore, it is more affordable and does not emit any greenhouse gases. This review compares conventional desalination plants with nuclear-powered desalination plants in terms of cost, energy consumption, water recovery, and environmental issues. On the basis of the review conducted, nuclear desalination has been demonstrated to be technically feasible and economically competitive with a variety of fossil fuels, renewable energy sources, and waste heat sources. Nuclear sources have been criticized due to their lack of safety. But studies show, if we were able to handle the issue with care, the problems could be eliminated. Here we're looking at the Seawater Reverse Osmosis Plant (SWROP) at Kudankulam Nuclear Power Plant in Tamil Nadu, India and review the further possibility of implementing nuclear desalination technology in other states of India.

Keywords: energy consumption, environmental impacts, nuclear desalination, water recovery

Procedia PDF Downloads 211

Authors: Denise Levy, Helen J. Khoury

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Nuclear technology is a controversial issue among a great share of the Brazilian population. Misinformation and common wrong beliefs confuse public’s perceptions and the scientific community is expected to offer a wider perspective on the benefits and risks resulting from ionizing radiation in everyday life. Attentive to the need of new approaches between science and society, the Nuclear Energy Museum, in northeast Brazil, is an initiative created to communicate the growing impact of the beneficial applications of nuclear technology in medicine, industry, agriculture and electric power generation. Providing accessible scientific information, the museum offers a rich learning environment, making use of different educational strategies, such as films, interactive panels and multimedia learning tools, which not only increase the enjoyment of visitors, but also maximize their learning potential. Developed according to modern active learning instructional strategies, multimedia materials are designed to present the increasingly role of nuclear science in modern life, transforming science education into a meaningful learning experience. In year 2016, nine different interactive computer-based activities were developed, presenting curiosities about ionizing radiation in different landmarks around the world, such as radiocarbon dating works in Egypt, nuclear power generation in France and X-radiography of famous paintings in Italy. Feedback surveys have reported a high level of visitors’ satisfaction, proving the high quality experience in learning nuclear science at the museum. The Nuclear Energy Museum is the first and, up to the present time, the only permanent museum in Brazil devoted entirely to nuclear science.

Keywords: nuclear technology, multimedia learning tools, science museum, society and education

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Authors: Jun Su Ha

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Various advanced technologies will be adopted in Advanced Control Rooms (ACRs) of advanced Nuclear Power Plants (NPPs), which is thought to increase operators’ performance. However, potential human factors issues coupled with digital technologies might be troublesome. Human factors issues in ACRs are identified and strategies (or countermeasures) for evaluating and analyzing each of issues are addressed in this study.

Keywords: advanced control room, human factor issues, human performance, human error, nuclear power plant

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Authors: Moon Byung Woo, Seok Chang-Sung, Koo Jae-Mean, Kim Sang-Young, Choi Jae Gu, Huh Nam-Su

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Recently, the occurrence of the earthquake has increased sharply and many of the casualties have occurred worldwide, due to the influence of earthquakes. Especially, the Fukushima nuclear power plant accident which was caused by the earthquake in 2011 has significantly increased the fear of people and the demand for the safety of the nuclear power plant. Thus, in order to prevent the earthquake accident at nuclear power plant, it is important to evaluate the fracture toughness considering the seismic loading rate. To obtain fracture toughness for the safety evaluation of nuclear power plant, it is desirable to perform experiments with a real scale pipe which is expensive and hard to perform. Therefore, many researchers have proposed various test specimens to replicate the fracture toughness of a real scale pipe. Since such specimens have several problems, the equivalent stress gradient (ESG) specimen has been recently suggested. In this study, in order to consider the effects of the dynamic loading speed on fracture toughness, the experiment was conducted by applying five different kinds of test speeds using an ESG specimen. In addition, after we performed the fracture toughness test under dynamic loading with different speeds using an ESG specimen and a standard specimen, we compared them with the test results under static loading.

Keywords: dynamic loading speed, fracture toughness, load-ratio-method, equivalent stress gradient (ESG) specimen

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Authors: Dong Wook Lee

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This paper presents a method to analyze the stiffness of the expansion joint with structural support using a hybrid method combining computational and analytical methods. Many expansion joints found in tubes and ducts of mechanical structures are designed to absorb thermal expansion mismatch between their structural members and deal with misalignments introduced from the assembly/manufacturing processes. One of the important design perspectives is the system’s vibrational characteristics. We calculate the stiffness as a characterization parameter for structural joint systems using a combined Finite Element Analysis (FEA) and an analytical method. We apply the methods to two sample applications: external configurations of aircraft engines and nuclear power plant structures.

Keywords: expansion joint, expansion joint stiffness, finite element analysis, nuclear power plants, aircraft engine external configurations

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Authors: S. W. Chen, W. K. Lin, J. R. Wang, C. Shih, H. T. Lin, H. C. Chang, W. Y. Li

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Kuosheng nuclear power plant (NPP) is a BWR/6 type NPP and located on the northern coast of Taiwan. First, Kuosheng NPP TRACE model were developed in this research. In order to assess the system response of Kuosheng NPP TRACE model, startup tests data were used to evaluate Kuosheng NPP TRACE model. Second, the over pressurization transient analysis of Kuosheng NPP TRACE model was performed. Besides, in order to confirm the mechanical property and integrity of fuel rods, FRAPTRAN analysis was also performed in this study.

Keywords: TRACE, safety analysis, BWR/6, FRAPTRA

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Authors: Akbar Rahmani Nejad, Pejman Rahmani Nejad, Ahmad Rahmani Nejad

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we proposed the arresting of atmospheric lightning and passing the electrical current of lightning-bolts through underground water tanks to produce Hydrogen and restoring Hydrogen in reservoirs to be used later as clean and sustainable energy. It is proposed to implement this method for storage of extra electrical power (instead of lightning energy) during low energy demand periods to produce hydrogen as a clean energy source to store in big reservoirs and later generate electricity by burning the stored hydrogen at an appropriate time. This method prevents the complicated process of changing the output power of nuclear power plants. It is possible to pass an electric current through sodium chloride solution to produce chlorine and sodium or human waste to produce Methane, etc. however atmospheric lightning is an accidental phenomenon, but using this free energy just by connecting the output of lightning arresters to the output of power plant during low energy demand period which there is no significant change in the design of power plant or have no cost, can be considered completely an economical design

Keywords: hydrogen gas, lightning energy, power plant, resistive element

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Authors: Phanindra Prasad Thummala, Umran Tezcan Un

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In this study, hydrodynamics analysis of two batch type electrocoagulation stirred tank reactors, used for the electrocoagulation treatment of Cr(VI) wastewater, was carried using computational fluid dynamics (CFD). The aim of the study was to evaluate the impact of mixing characteristics on overall performance of electrocoagulation reactor. The CFD simulations were performed using ANSYS FLUENT 14.4 software. The mixing performance of each reactor was evaluated by numerically modelling tracer dispersion in each reactor configuration. The uniformity in tracer dispersion was assumed when 90% of the ratio of the maximum to minimum concentration of the tracer was realized. In parallel, experimental evaluation of both the electrocoagulation reactors for removal of Cr(VI) from wastewater was also carried out. The results of CFD and experimental analysis clearly show that the reactor which can give higher uniformity in lesser time, will perform better as an electrocoagulation reactor for removal of Cr(VI) from wastewater.

Keywords: CFD, stirred tank reactors, electrocoagulation, Cr(VI) wastewater

Procedia PDF Downloads 462

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

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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: Danying Gu, Xiaoyan Li, Yuanlei He

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As a knowledge-intensive industry, nuclear industry highly values the importance of safety and quality. The life cycle of a NPP (Nuclear Power Plant) can last 100 years from the initial research and design to its decommissioning. How to implement the high-quality knowledge management and how to contribute to a more safe, advanced and economic NPP (Nuclear Power Plant) is the most important issue and responsibility for knowledge management. As the lead of nuclear industry, nuclear research and design institute has competitive advantages of its advanced technology, knowledge and information, DKM (Design Knowledge Management) of nuclear research and design institute is the core of the knowledge management in the whole nuclear industry. In this paper, the study and practice on DKM and knowledge transfer across the life cycle of a new-built NPP in China is introduced. For this digital intelligent NPP, the whole design process is based on a digital design platform which includes NPP engineering and design dynamic analyzer, visualization engineering verification platform, digital operation maintenance support platform and digital equipment design, manufacture integrated collaborative platform. In order to make all the design data and information transfer across design, construction, commissioning and operation, the overall architecture of new-built digital NPP should become a modern knowledge management system. So a digital information transfer model across the NPP life cycle is proposed in this paper. The challenges related to design knowledge transfer is also discussed, such as digital information handover, data center and data sorting, unified data coding system. On the other hand, effective delivery of design information during the construction and operation phase will contribute to the comprehensive understanding of design ideas and components and systems for the construction contractor and operation unit, largely increasing the safety, quality and economic benefits during the life cycle. The operation and maintenance records generated from the NPP operation process have great significance for maintaining the operating state of NPP, especially the comprehensiveness, validity and traceability of the records. So the requirements of an online monitoring and smart diagnosis system of NPP is also proposed, to help utility-owners to improve the safety and efficiency.

Keywords: design knowledge management, digital nuclear power plant, knowledge transfer, life cycle

Procedia PDF Downloads 273

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

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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

Authors: Avtandil Okrostsvaridze, Salome Gogoladze

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In the light of exhaustion of hydrocarbon reserves of new energy resources, its search is of vital importance problem for the modern civilization. At the time of energy resource crisis, the radioactive element thorium (232Th) is considered as the main energy resource for the future of our civilization. Modern industry uses thorium in high-temperature and high-tech tools, but the most important property of thorium is that like uranium it can be used as fuel in nuclear reactors. However, thorium has a number of advantages compared to this element: Its concentration in the earth crust is 4-5 times higher than uranium; extraction and enrichment of thorium is much cheaper than of uranium; it is less radioactive; its waste products complete destruction is possible; thorium yields much more energy than uranium. Nowadays, developed countries, among them India and China, have started intensive work for creation of thorium nuclear reactors and intensive search for thorium reserves. It is not excluded that in the next 10 years these reactors will completely replace uranium reactors. Thorium ore mineralization is genetically related to alkaline-acidic magmatism. Thorium accumulations occur as in endogen marked as in exogenous conditions. Unfortunately, little is known about the reserves of this element in Georgia, as planned prospecting-exploration works of thorium have never been carried out here. Although, 3 ore occurrences of this element are detected: 1) In the Greater Caucasus Kakheti segment, in the hydrothermally altered rocks of the Lower Jurassic clay-shales, where thorium concentrations varied between 51 - 3882g/t; 2) In the eastern periphery of the Dzirula massif, in the hydrothermally alteration rocks of the cambrian quartz-diorite gneisses, where thorium concentrations varied between 117-266 g/t; 3) In active contact zone of the Eocene volcanites and syenitic intrusive in Vakijvari ore field of the Guria region, where thorium concentrations varied between 185 – 428 g/t. In addition, geological settings of the areas, where thorium occurrences were fixed, give a theoretical basis on possible accumulation of practical importance thorium ores. Besides, the Black Sea Guria region magnetite sand which is transported from Vakijvari ore field, should contain significant reserves of thorium. As the research shows, monazite (thorium containing mineral) is involved in magnetite in the form of the thinnest inclusions. The world class thorium deposit concentrations of this element vary within the limits of 50-200 g/t. Accordingly, on the basis of these data, thorium resources found in Georgia should be considered as perspective ore deposits. Generally, we consider that complex investigation of thorium should be included into the sphere of strategic interests of the state, because future energy of Georgia, will probably be thorium.

Keywords: future energy, Georgia, ore field, thorium

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Authors: J. R. Wang, H. T. Lin, H. C. Chen, C. Shih, S. W. Chen, S. C. Chiang, C. C. Liu

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In this research, TRACE model of Chinshan BWR/4 Nuclear Power Plant (NPP) has been developed for the simulation and analysis of Ultimate Response Guideline (URG). The main actions of URG are the depressurization and low pressure water injection of reactor and containment venting. This research focuses to verify the URG efficiency under Fukushima-like conditions. Trace analysis results show that the URG can keep the PCT below the criteria 1088.7 K under Fukushima-like conditions. It indicated that Chinshan NPP was safe.

Keywords: BWR, trace, safety analysis, URG

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Authors: Onur Yuksel

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In today’s complex foreign relations between countries, the nuclear enrichment and nuclear weapon have become a threat for all states in the world. There are couple isolated states which have capacity to produce nuclear weapons such as Iran and North Korea. In this article, Iran nuclear negotiation was analyzed in terms of its relations especially with The United States in order to find the important factors that affect the course of the ongoing nuclear negotiation. In this sense, the Force Field Analysis was used by determining and setting forth Driving and Restraining Forces of the nuclear negotiations in order to see the big picture and to develop strategies that may improve the long-term ongoing Iran nuclear negotiations. It is found that Iran nuclear negotiation heavily depends on breaking down the idea of Iran’s supporting terrorist organizations and being more transparent about nuclear and uranium enrichment. Also, it was found that Iran has to rebuild its relations with Western countries, especially with the United States. In addition, the counties— who contribute to Iran nuclear negotiations— will need to work on the dynamics and drivers of the Israel and Iran relations in order to peacefully transform the conflict between the two states.

Keywords: driving force, Iran nuclear negotiation, restraining force, the force field analysis

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Authors: Uchenna Egwu, Paul Jonathan Sallis

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Perennial ryegrass is an abundant renewable lignocellulosic biofuel feedstock for biomethane production through anaerobic digestion (AD). In this study, six anaerobic continuously stirred tank reactors (CSTRs) were set up in three pairs. Each pair of the CSTRs was then used to study the effects of operating temperatures – psychrophilic, mesophilic, and thermophilic, and optimum pH on the specific methane production (SMP) of the ryegrass during AD under discontinuous daily feeding conditions. The reactors were fed at an organic loading rate (OLR) ranging from 1-1.5 kgVS.L⁻¹d⁻¹ and hydraulic residence time, HRT=20 days for 140 days. The pH of the digesters was maintained at the range of 6.8-7.2 using 1 M NH₄HCO₃ solution, but this was replaced with biomass ash-extracts from day 105-140. The results obtained showed that the mean SMP of ryegrass measured between HRT 3 and 4 were 318.4, 425.4 and 335 N L CH₄ kg⁻¹VS.d⁻¹ for the psychrophilic (25 ± 2°C), mesophilic (40 ± 1°C) and thermophilic (60 ± 1°C) temperatures respectively. It was also observed that the buffering ability of the reactors increased with operating temperature, probably due to an increase in the solubility of ammonium bicarbonate (NH₄HCO₃) with temperature. The reactors also achieved a mean VS destruction of 61.9, 68.5 and 63.5%, respectively, which signifies that the mesophilic reactors achieved the highest specific methane production (SMP), while the psychrophilic reactors achieved the lowest. None of the reactors attained steady-state condition due to the discontinuous daily feeding times, and therefore, such feeding practice may not be the most effective for maximum biogas production over long periods of time. The addition of NH₄HCO₃ as supplement provided a good buffering condition in these AD digesters, but the digesters failed in the long run due to inhibition from the accumulation of free ammonia, which later led to decrease in pH, acidification, and souring of the digesters. However, the addition of biomass ash extracts was shown to potentially revive failed AD reactors by providing an adequate buffering and essential trace nutrient supplements necessary for optimal bacterial growth.

Keywords: anaerobic digestion, discontinuous feeding, perennial ryegrass, specific methane production, supplements, temperature

Procedia PDF Downloads 127

Authors: Shivani Gupta, Shrishti Chaturvedi

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By using a miniscule of nuclear matter, the problem of immeasurable human needs for energy can be resolved. However since nuclear energy also has the inherent potential for catastrophic destruction, one should be extremely mindful of the consequences should a mischance occur. Civil Nuclear Liability has recently gained a lot of momentum after India entered into agreements with nations like United States of America, France and others. Also now India is a part of the Convention on Supplementary Compensation (CSC). With a history of Bhopal Gas Tragedy, India is now much more vigilant about the latest developments in this sector. Therefore, it has become imperative to analyses the liability regime in the background of international conventions such as Vienna Convention 1963, Paris Convention 1960, Convention on Supplementary Compensation, 1997 and others. Also the present Indian legal scenarios in this regard which are derived from Civil Liability for Nuclear Damages Act, 2010 and Civil Liability for Nuclear Damages Rules, 2011 have also been extensively discussed in the paper.

Keywords: nuclear liability, civil liability for nuclear damages act, 2010, civil liability for nuclear damages rules, India

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Authors: D. Abdelrazek, M. NaguibAly, A. A. Badawi, Asmaa G. Abo Elnour, A. A. El-Kafas

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This work is developed within IAEA Coordinated Research Program 1496, “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors.” The study investigates the capability of Code RELAP5/Mod3.4 to solve complex geometry complexity. Its results are compared to the results of PARET, a common code in thermal hydraulic analysis for research reactors, belonging to MTR-PC groups. The WWR-SM reactor at the Institute of Nuclear Physics (INP) in the Republic of Uzbekistan is simulated using both PARET and RELAP5 at steady state. Results from the two codes are compared. REALP5 code succeeded in solving the complex fuel geometry. The PARET code needed some calculations to obtain the final result. Although the final results from the PARET are more accurate, the small differences in both results makes using RELAP5 code recommended in case of complex fuel assemblies.

Keywords: complex fuel geometry, PARET, RELAP5, WWR-SM reactor

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Authors: M. Gomaa Abdoelatef, Robert M. Field, Lee, Yong-Kwan

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Growing human populations have placed increased demands on water supplies and a heightened interest in desalination infrastructure. Key elements of the economics of desalination projects are thermal and electrical inputs. With growing concerns over the use of fossil fuels to (indirectly) supply these inputs, coupling of desalination with nuclear power production represents a significant opportunity. Individually, nuclear and desalination technologies have a long history and are relatively mature. For desalination, Reverse Osmosis (RO) has the lowest energy inputs. However, the economically driven output quality of the water produced using RO, which uses only electrical inputs, is lower than the output water quality from thermal desalination plants. Therefore, modern desalination projects consider that RO should be coupled with thermal desalination technologies (MSF, MED, or MED-TVC) with attendant steam inputs to permit blending to produce various qualities of water. A large nuclear facility is well positioned to dispatch large quantities of both electrical and thermal power. This paper considers the supply of thermal energy to a large desalination facility to examine heat balance impact on the nuclear steam cycle. The APR1400 nuclear plant is selected as prototypical from both a capacity and turbine cycle heat balance perspective to examine steam supply and the impact on electrical output. Extraction points and quantities of steam are considered parametrically along with various types of thermal desalination technologies to form the basis for further evaluations of economically optimal approaches to the interface of nuclear power production with desalination projects. In our study, the thermodynamic evaluation will be executed by DE-TOP which is the IAEA desalination program, it is approved to be capable of analyzing power generation systems coupled to desalination systems through various steam extraction positions, taking into consideration the isolation loop between the APR-1400 and the thermal desalination plant for safety concern.

Keywords: APR-1400, desalination, DE-TOP, IAEA, MSF, MED, MED-TVC, RO

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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: Altab Hossain, Shakerul Islam, Golamur R. Khan, Abu Zafar M. Salahuddin

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An effective identification of breakdowns is of premier importance for the safe and reliable operation of Nuclear Power Plants (NPP) and its associated facilities. A great number of monitoring and diagnosis methodologies are applied and used worldwide in areas such as industry, automobiles, hospitals, and power plant to detect and reduce human disasters. The potential consequences of several hazardous activities may harm the society using nuclear and its associated facilities. Hence, one of the most popular and effective methods to ensure safety and monitor the entire nuclear facility and imply risk-free operation without human interference during the hazardous situation is using a robot. Therefore, in this study, an advanced autonomous robot has been designed and developed that can monitor several parameters in the NPP to ensure the safety and do some risky job in case of nuclear disaster. The robot consisted of autonomous track following unit, data processing and transmitting unit can follow a straight line and take turn as the bank greater than 90 degrees. The developed robot can analyze various parameters such as temperature, altitude, radiation, obstacle, humidity, detecting fire, measuring distance, ultrasonic scan and taking the heat of any particular object. It has an ability to broadcast live stream and can record the document to its own server memory. There is a separate control unit constructed with a baseboard which processes the recorded data and a transmitter which transmits the processed data. To make the robot user-friendly, the code is developed such a way that a user can control any of robotic arm as per types of work. To control at any place and without the track, there is an advanced code has been developed to take manual overwrite. Through this process, administrator who has logged in permission to Dynamic Host Client Protocol (DHCP) can make the handover of the control of the robot. In this process, this robot is provided maximum nuclear security from being hacked. Not only NPP, this robot can be used to maximize the real-time monitoring system of any nuclear facility as well as nuclear material transportation and decomposition system.

Keywords: nuclear power plant, radiation, dynamic host client protocol, nuclear security

Procedia PDF Downloads 209

Authors: J. R. Wang, S. W. Chen, Y. Chiang, W. S. Hsu, J. H. Yang, Y. S. Tseng, C. Shih

Abstract:

In this research, the HABIT analysis methodology was established for Maanshan nuclear power plant (NPP). The Final Safety Analysis Report (FSAR), reports, and other data were used in this study. To evaluate the control room habitability under the CO₂ storage burst, the HABIT methodology was used to perform this analysis. The HABIT result was below the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the parameters (wind speed, atmospheric stability classification, air temperature, and control room intake flow rate) was also performed in this research.

Keywords: PWR, HABIT, Habitability, Maanshan

Procedia PDF Downloads 445

Authors: Y. Chiang, W. Y. Li, J. R. Wang, S. W. Chen, W. S. Hsu, J. H. Yang, Y. S. Tseng, C. Shih

Abstract:

In this research, the HABIT code was used to estimate the concentration under the CO₂ and H₂SO₄ storage burst conditions for Kuosheng nuclear power plant (NPP). The Final Safety Analysis Report (FSAR) and reports were used in this research. In addition, to evaluate the control room habitability for these cases, the HABIT analysis results were compared with the R.G. 1.78 failure criteria. The comparison results show that the HABIT results are below the criteria. Additionally, some sensitivity studies (stability classification, wind speed and control room intake rate) were performed in this study.

Keywords: BWR, HABIT, habitability, Kuosheng

Procedia PDF Downloads 489

Authors: J. R. Wang, Y. Chiang, W. Y. Li, H. T. Lin, H. C. Chen, C. Shih, S. W. Chen

Abstract:

The dry-storage systems of nuclear power plants (NPPs) in Taiwan have become one of the major safety concerns. There are two steps considered in this study. The first step is the verification of the TRACE by using VSC-17 experimental data. The results of TRACE were similar to the VSC-17 data. It indicates that TRACE has the respectable accuracy in the simulation and analysis of the dry-storage systems. The next step is the application of TRACE in the dry-storage system of Kuosheng NPP (BWR/6). Kuosheng NPP is the second BWR NPP of Taiwan Power Company. In order to solve the storage of the spent fuels, Taiwan Power Company developed the new dry-storage system for Kuosheng NPP. In this step, the dry-storage system model of Kuosheng NPP was established by TRACE. Then, the steady state simulation of this model was performed and the results of TRACE were compared with the Kuosheng NPP data. Finally, this model was used to perform the safety analysis of Kuosheng NPP dry-storage system. Besides, FRAPTRAN was used tocalculate the transient performance of fuel rods.

Keywords: BWR, TRACE, FRAPTRAN, dry-storage

Procedia PDF Downloads 519

Authors: Muhammad Jawad Hashmi

Abstract:

The idea of nuclear terrorism is as old as nuclear weapons but the global concerns of likelihood of nuclear terrorism are uncertain. Post 9/11 trends manifest that terrorists are believers of massive causalities. Innovation in terrorist’s tactics, sophisticated weaponry, vulnerability, theft and smuggling of nuclear/radiological material, connections between terrorists, black market and rough regimes are signaling seriousness of upcoming challenges as well as global trends of “terror-transnationalism.” Furthermore, the International-Atomic-Energy-Agency’s database recorded 2734 incidents regarding misuse, unauthorized possession, trafficking of nuclear material etc. Since, this data also includes incidents from south Asia, so, there is every possibility to claim that such illicit activities may increase in future, mainly due to expansion of nuclear industry in South Asia. Moreover, due to such mishaps the region is vulnerable to threats of nuclear terrorism. This is also a reason that the region is in limelight along with issues such as rapidly growing nuclear arsenals, nuclear safety and security, terrorism and political instability. With this backdrop, this study is aimed to investigate the prevailing threats and challenges in South Asia vis a vis nuclear safety and security. A comparative analysis of the overall capabilities would be done to identify the areas of cooperation to eliminate the probability of nuclear/radiological terrorism in the region.

Keywords: nuclear terrorism, safety, security, South Asia, india, Pakistan

Procedia PDF Downloads 356