Search results for: nuclear fuel fission

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

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Fuel rod analysis program transient (FRAPTRAN) code was used to study the fuel rod performance during a postulated large break loss of coolant accident (LBLOCA) in Maanshan nuclear power plant (NPP). Previous transient results from thermal hydraulic code, TRACE, with the same LBLOCA scenario, were used as input boundary conditions for FRAPTRAN. The simulation results showed that the peak cladding temperatures and the fuel center line temperatures were all below the 10CFR50.46 LOCA criteria. In addition, the maximum hoop stress was 18 MPa and the oxide thickness was 0.003 mm for the present simulation cases, which are all within the safety operation ranges. The present study confirms that this analysis method, the FRAPTRAN code combined with TRACE, is an appropriate approach to predict the fuel integrity under LBLOCA with operational ECCS.

Keywords: FRAPTRAN, TRACE, LOCA, PWR

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Authors: Adem Acır

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In this study, time dependent neutronic analysis of incineration of minor actinides of a Laser Fusion Inertial Confinement Fusion Fission Energy (LIFE) engine was performed. The calculations were carried out by using MCNP codes with ENDF/B.VI neutron data library. In the neutronic calculations, TRISO particles fueled with minor actinides with natural lithium coolant were performed. The natural lithium cooled LIFE engine used 10 % TRISO fuel minor actinides composition. Tritium breeding ratios (TBR) and energy multiplication factor (M) burnup values were computed as 1.46 and 3.75, respectively. The reactor operation time was calculated as ~ 21 years. The burnup values were obtained as ~1060 GWD/MT, respectively. As a result, the very higher burnup were achieved of LIFE engine.

Keywords: Monte Carlo, minor actinides, nuclear waste, LIFE engine

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Authors: Djalal Hamed

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The aim of this paper is to perform, by mean of the finite volume method, a numerical solution of the transient natural convection in a narrow rectangular channel between two vertical parallel Material Testing Reactor (MTR)-type fuel plates, imposed under a heat flux with a cosine shape to determine the margin of the nuclear core power at which the natural convection cooling mode can ensure a safe core cooling, where the cladding temperature should not reach a specific safety limits (90 °C). For this purpose, a computer program is developed to determine the principal parameters related to the nuclear core safety, such as the temperature distribution in the fuel plate and in the coolant (light water) as a function of the reactor core power. Throughout the obtained results, we noticed that the core power should not reach 400 kW, to ensure a safe passive residual heat removing from the nuclear core by the upward natural convection cooling mode.

Keywords: buoyancy force, friction force, finite volume method, transient natural convection

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Authors: Dmitry V. Fomichev, Vladimir V. Solonin

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This paper presents the findings from a numerical simulation of the flow in 37-rod fuel assembly models spaced by a double-wire trapezoidal wrapping as applied to the BREST-OD-300 experimental nuclear reactor. Data on a high static pressure distribution within the models, and equations for determining the fuel bundle flow friction factors have been obtained. Recommendations are provided on using the closing turbulence models available in the ANSYS Fluent. A comparative analysis has been performed against the existing empirical equations for determining the flow friction factors. The calculated and experimental data fit has been shown. An analysis into the experimental data and results of the numerical simulation of the BREST-OD-300 fuel rod assembly hydrodynamic performance are presented.

Keywords: BREST-OD-300, ware-spaces, fuel assembly, computation fluid dynamics

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

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

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

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Authors: Dimitrie Marinceu, Alan Murchison

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The Canadian Used Fuel Container (UFC) is a mid-size hemispherical headed copper coated steel container measuring 2.5 meters in length and 0.5 meters in diameter containing 48 used fuel bundles. The contained used fuel produces significant gamma radiation requiring automated assembly processes to complete the assembly. The design throughput of 2,500 UFCs per year places constraints on equipment and hot cell design for repeatability, speed of processing, robustness and recovery from upset conditions. After UFC assembly, the UFC is inserted into a Buffer Box (BB). The BB is made from adequately pre-shaped blocks (lower and upper block) and Highly Compacted Bentonite (HCB) material. The blocks are practically ‘sandwiching’ the UFC between them after assembly. This paper identifies one possible approach for the BB automatic assembly cell and processes. Automation of the BB assembly will have a significant positive impact on nuclear safety, quality, productivity, and reliability.

Keywords: used fuel packing plant, automatic assembly cell, used fuel container, buffer box, deep geological repository

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Authors: Terry Lucas

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Calculations are performed to quantify the potential contribution of each greenhouse gas emission reduction strategy. This approach facilitates the visualisation of the relative benefits of each, and it provides a potential baseline for the development of a plan of action that is rooted in quantitative evaluation. Emissions reductions are converted to potential de-escalation of global average temperature. A comprehensive plan is then presented which shows the potential benefits all the way out to year 2100. A target temperature de-escalation of 2oC was selected, but the plan shows a benefit of only 1.225oC. This latter disappointing result is in spite of new and powerful technologies introduced into the equation. These include nuclear fusion and alternative nuclear fission processes. Current technologies such as wind, solar and electric vehicles show surprisingly small constributions to the whole.

Keywords: climate change, emissions, drawdown, energy

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

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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

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Authors: Djalal Hamed

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The aim of this paper is to perform an analytic solution of steady state natural convection in a narrow rectangular channel between two vertical parallel MTR-type fuel plates, imposed under a cosine shape heat flux to determine the margin of the nuclear core power at which the natural convection cooling mode can ensure a safe core cooling, where the cladding temperature should not be reach the specific safety limits (90 °C). For this purpose, a simple computer program is developed to determine the principal parameter related to the nuclear core safety such as the temperature distribution in the fuel plate and in the coolant (light water) as a function of the reactor power. Our results are validated throughout a comparison against the results of another published work, which is considered like a reference of this study.

Keywords: buoyancy force, friction force, natural convection, thermal hydraulic analysis, vertical heated rectangular channel

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Authors: A. Orlov, V. Rachkov

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The future of Nuclear Energetics is seen in fast reactors with inherent safety working in the closed nuclear fuel cycle. The concept of inherent safety, which lies in deterministic elimination of the most severe accidents due to inherent properties of the reactor rather than through building up engineered barriers, is a cornerstone of success in ensuring safety and economic efficiency of future Nuclear Energetics. The focus of this paper is one of the key elements of inherent safety - the lead coolant of a nuclear reactor. Advantages of lead coolant for reactor application, influence on safety are reviewed. BREST-OD-300 fast reactor, currently being developed in Russia withing the “Proryv” Project utilizes lead coolant and a special set of measures and devices, called technology of lead coolant that ensures safe operation in a wide range of temperatures. Here these technological elements are reviewed, and current progress in their development is discussed.

Keywords: BREST-OD-300. , fast reactor, inherent safety, lead coolant

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

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

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

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Authors: Mohamad Saab, Sidi Souvi

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In a major nuclear accident, the released fission products (FPs) and the structural materials are likely to influence the transport of iodine in the reactor coolant system (RCS) of a pressurized water reactor (PWR). So far, the thermodynamic data on cesium and silver species used to estimate the magnitude of FP release show some discrepancies, data are scarce and not reliable. For this reason, it is crucial to review the thermodynamic values related to cesium and silver materials. To this end, we have used state-of-the-art quantum chemical methods to compute the formation enthalpies and entropies of AgHMoO₄, CsHMoO₄, and AgCsMoO₄ in the gas phase. Different quantum chemical methods have been investigated (DFT and CCSD(T)) in order to predict the geometrical parameters and the energetics including the correlation energy. The geometries were optimized with TPSSh-5%HF method, followed by a single point calculation of the total electronic energies using the CCSD(T) wave function method. We thus propose with a final uncertainty of about 2 kJmol⁻¹ standard enthalpies of formation of AgHMoO₄, CsHMoO₄, and AgCsMoO₄.

Keywords: nuclear accident, ASTEC code, thermochemical database, quantum chemical methods

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

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

Keywords: particulate matter, cardiomyocyte, apoptosis, mitochondria

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Authors: Xingyu Peng, Qingyuan Hu, Xuebin Zhu, Xi Yuan

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Neutron spectrometry has contributed much to the development of nuclear physics since 1932 and has also become an importance tool in several other fields, notably nuclear technology, fusion plasma diagnostics and radiation protection. Compared with neutron fluxes, neutron spectra can provide more detailed information on the internal physical process of neutron sources, such as fast neutron reactors, fusion plasma, fission-fusion hybrid reactors, and so on. However, high performance neutron spectrometer is not so commonly available as it requires the use of large and complex instrumentation. This work describes the development and characterization of a compact magnetic proton recoil (MPR) spectrometer for high-resolution measurements of fast neutron spectra. The compact MPR spectrometer is featured by its large recoil angle, small size permanent analysis magnet, short beam transport line and dual-purpose detector array for both steady state and pulsed neutron spectra measurement. A 3-dimensional electromagnetic particle transport code is developed to simulate the response function of the spectrometer. Simulation results illustrate that the performance of the spectrometer is mainly determined by n-p recoil foil and proton apertures, and an overall energy resolution of 3% is achieved for 14 MeV neutrons. Dedicated experiments using alpha source and mono-energetic neutron beam are employed to verify the simulated response function of the compact MPR spectrometer. These experimental results show a good agreement with the simulated ones, which indicates that the simulation code possesses good accuracy and reliability. The compact MPR spectrometer described in this work is a valuable tool for fast neutron spectra measurements for the fission or fusion devices.

Keywords: neutron spectrometry, magnetic proton recoil spectrometer, neutron spectra, fast neutron

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

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

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

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Authors: Aydin Azizi, Abdurrahman Tanira

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With the onset of climate change and limited fuel resources, improving fuel efficiency has become an important part of the motor industry. To maximize fuel efficiency, development of technologies must come hand-in-hand with awareness of efficient driving strategies. This study aims to explore the various driving habits that can impact fuel efficiency by reviewing available literature. Such habits include sudden and unnecessary acceleration or deceleration, improper hardware maintenance, driving above or below optimum speed and idling. By studying such habits and ultimately applying it to driving techniques, in combination with improved mechanics of the car, will optimize the use of fuel.

Keywords: fuel efficiency, driving techniques, optimum speed, optimizing fuel consumption

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Authors: Sivakumar Kottapalli, Abdesselam Abdelouas, Christoph Hartnack

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Spent nuclear fuel generates a mixed field of ionizing radiation to the water. This radiation field is generally dominated by gamma rays and a limited flux of fast neutrons. The fuel cladding effectively attenuates beta and alpha particle radiation. Small fraction of the spent nuclear fuel exhibits some degree of fuel cladding penetration due to pitting corrosion and mechanical failure. Breaches in the fuel cladding allow the exposure of small volumes of water in the cask to alpha and beta ionizing radiation. The safety of the transport of radioactive material is assured by the package complying with the IAEA Requirements for the Safe Transport of Radioactive Material SSR-6. It is of high interest to avoid generation of hydrogen inside the cavity which may to an explosive mixture. The risk of hydrogen production along with other radiation gases should be analyzed for a typical spent fuel for safety issues. This work aims to perform a realistic study of the production of hydrogen by radiolysis assuming most penalizing initial conditions. It consists in the calculation of the radionuclide inventory of a pellet taking into account the burn up and decays. Westinghouse 17X17 PWR fuel has been chosen and data has been analyzed for different sets of enrichment, burnup, cycles of irradiation and storage conditions. The inventory is calculated as the entry point for the simulation studies of hydrogen production by radiolysis kinetic models by MAKSIMA-CHEMIST. Dose rates decrease strongly within ~45 μm from the fuel surface towards the solution(water) in case of alpha radiation, while the dose rate decrease is lower in case of beta and even slower in case of gamma radiation. Calculations are carried out to obtain spectra as a function of time. Radiation dose rate profiles are taken as the input data for the iterative calculations. Hydrogen yield has been found to be around 0.02 mol/L. Calculations have been performed for a realistic scenario considering a capsule containing the spent fuel rod. Thus, hydrogen yield has been debated. Experiments are under progress to validate the hydrogen production rate using cyclotron at > 5MeV (at ARRONAX, Nantes).

Keywords: radiolysis, spent fuel, hydrogen, cyclotron

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Authors: Guanyu Jiang, Donghai Xu, Huanteng Liu

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After the Fukushima-Daiichi accident, the development of accident tolerant fuel cladding materials to improve reactor safety has become a hot topic in the field of nuclear industry. ZrO₂ has a satisfactory neutron economy and can guarantee the fission chain reaction process, which enables it to be a promising coating for zirconium alloy cladding. Maintaining a good corrosion resistance in primary coolant loop during normal operations of Pressurized Water Reactors is a prerequisite for ZrO₂ as a protective coating on zirconium alloy cladding. Research on the corrosion performance of ZrO₂ coating in nuclear water chemistry is relatively scarce, and existing reports failed to provide an in-depth explanation for the failure causes of ZrO₂ coating. Herein, a detailed corrosion process of ZrO₂ coating in lithiated water at 360 °C and 18.5 MPa was proposed based on experimental research and molecular dynamics simulation. Lithiated water with different LiOH solutions in the present work was deaerated and had a dissolved oxygen concentration of < 10 ppb. The concentration of Li (as LiOH) was determined to be 2.3 ppm, 70 ppm, and 500 ppm, respectively. Corrosion tests were conducted in a static autoclave. Modeling and corresponding calculations were operated on Materials Studio software. The calculation of adsorption energy and dynamics parameters were undertaken by the Energy task and Dynamics task of the Forcite module, respectively. The protective effect and failure mechanism of ZrO₂ coating on Zry-4 under varied LiOH concentrations was further revealed by comparison with the coating corrosion performance in pure water (namely 0 ppm Li). ZrO₂ coating provided a favorable corrosion protection with the occurrence of localized corrosion at low LiOH concentrations. Factors influencing corrosion resistance mainly include pitting corrosion extension, enhanced Li+ permeation, short-circuit diffusion of O²⁻ and ZrO₂ phase transformation. In highly-concentrated LiOH solutions, intergranular corrosion, internal oxidation, and perforation resulted in coating failure. Zr ions were released to coating surface to form flocculent ZrO₂ and ZrO₂ clusters due to the strong diffusion and dissolution tendency of α-Zr in the Zry-4 substrate. Considering that primary water of Pressurized Water Reactors usually includes 2.3 ppm Li, the stability of ZrO₂ make itself a candidate fuel cladding coating material. Under unfavorable conditions with high Li concentrations, more boric acid should be added to alleviate caustic corrosion of ZrO₂ coating once it is used. This work can provide some references to understand the service behavior of nuclear coatings under variable water chemistry conditions and promote the in-pile application of ZrO₂ coating.

Keywords: ZrO₂ coating, Zry-4, corrosion behavior, failure mechanism, LiOH concentration

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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: Sewon Kim, Changyeop Lee

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A new concept of in-furnace partial oxidation combustion is successfully applied in this research. The burner is designed such that liquid fuel is prevaporized in the furnace then injected into a fuel rich combustion zone so that a partial oxidation reaction occurs. The effects of equivalence ratio, thermal load, injection distance and fuel distribution ratio on the NOx and CO are experimentally investigated. This newly developed burner showed very low NOx emission level, about 15 ppm when light oil is used as a fuel.

Keywords: burner, low NOx, liquid fuel, partial oxidation

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Authors: Debalina Ghoshal

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Nuclear security is the ‘prevention and detection of, and response to unauthorised removal, sabotage, unauthorised access, illegal transfer or other malicious acts involving nuclear or radiological material or their associated facilities.’ Ever since the end of Cold War, nuclear materials security has remained a concern for global security. However, with the increase in terrorist attacks not just in India especially, security of nuclear materials remains a priority. Therefore, India has made continued efforts to tighten its security on nuclear materials to prevent nuclear theft and radiological terrorism. Nuclear security is different from nuclear safety. Physical security is also a serious concern and India had been careful of the physical security of its nuclear materials. This is more so important since India is expanding its nuclear power capability to generate electricity for economic development. As India targets 60,000 MW of electricity production by 2030, it has a range of reactors to help it achieve its goal. These include indigenous Pressurised Heavy Water Reactors, now standardized at 700 MW per reactor Light Water Reactors, and the indigenous Fast Breeder Reactors that can generate more fuel for the future and enable the country to utilise its abundant thorium resource. Nuclear materials security can be enhanced through two important ways. One is through proliferation resistant technologies and diplomatic efforts to take non proliferation initiatives. The other is by developing technical means to prevent any leakage in nuclear materials in the hands of asymmetric organisations. New Delhi has already implemented IAEA Safeguards on their civilian nuclear installations. Moreover, the IAEA Additional Protocol has also been ratified by India in order to enhance its transparency of nuclear material and strengthen nuclear security. India is a party to the IAEA Conventions on Nuclear Safety and Security, and in particular the 1980 Convention on the Physical Protection of Nuclear Material and its amendment in 2005, Code of Conduct in Safety and Security of Radioactive Sources, 2006 which enables the country to provide for the highest international standards on nuclear and radiological safety and security. India's nuclear security approach is driven by five key components: Governance, Nuclear Security Practice and Culture, Institutions, Technology and International Cooperation. However, there is still scope for further improvements to strengthen nuclear materials and nuclear security. The NTI Report, ‘India’s improvement reflects its first contribution to the IAEA Nuclear Security Fund etc. in the future, India’s nuclear materials security conditions could be further improved by strengthening its laws and regulations for security and control of materials, particularly for control and accounting of materials, mitigating the insider threat, and for the physical security of materials during transport. India’s nuclear materials security conditions also remain adversely affected due to its continued increase in its quantities of nuclear material, and high levels of corruption among public officials.’ This paper would study briefly the progress made by India in nuclear and nuclear material security and the step ahead for India to further strengthen this.

Keywords: India, nuclear security, nuclear materials, non proliferation

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Authors: Mohamed Y. M. Mohsen

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The viability of utilising mixed-oxide fuel (MOX) ((U₀.₉, rgPu₀.₁) O₂) as an accident-tolerant fuel (ATF) has been thoroughly investigated. MOX fuel provides the best example of a nuclear waste recycling process. The MCNPX 2.7 code was used to determine the main neutronic features, especially the radial power distribution, to identify the hot channel on which the thermal-hydraulic (TH) study was performed. Based on the computational fluid dynamics technique, the simulation of the rod-centered thermal-hydraulic subchannel model was implemented using COMSOL Multiphysics. TH analysis was utilised to determine the axially and radially distributed temperatures of the fuel and cladding materials, as well as the departure from the nucleate boiling ratio (DNBR) along the coolant channel. COMSOL Multiphysics can simulate reality by coupling multiphysics, such as coupling between heat transfer and solid mechanics. The main solid structure parameters, such as the von Mises stress, volumetric strain, and displacement, were simulated using this coupling. When the neutronic, TH, and solid structure performances of UO₂ and ((U₀.₉, rgPu₀.₁) O₂) were compared, the results showed considerable improvement and an increase in safety margins with the use of ((U₀.₉, rgPu₀.₁) O₂).

Keywords: mixed-oxide, MCNPX, neutronic analysis, COMSOL-multiphysics, thermal-hydraulic, solid structure

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Authors: Yudhitya Kusumawati, Yuki Mitsuya, Tomooki Shiba, Mitsuru Uesaka

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After the Fukushima Daiichi nuclear power reactor incident, there are a lot of unaccountable nuclear fuel debris in the reactor core area, which is subject to safeguard and criticality safety. Before the actual precise analysis is performed, preliminary on-site screening and mapping of nuclear debris activity need to be performed to provide a reliable data on the nuclear debris mass-extraction planning. Through a collaboration project with Japan Atomic Energy Agency, an on-site nuclear debris screening system by using dual energy X-Ray inspection and neutron energy resonance analysis has been established. By using the compact and mobile pulsed neutron source constructed from 3.95 MeV X-Band electron linac, coupled with Tungsten as electron-to-photon converter and Beryllium as a photon-to-neutron converter, short-distance neutron Time of Flight measurement can be performed. Experiment result shows this system can measure neutron energy spectrum up to 100 eV range with only 2.5 meters Time of Flightpath in regards to the X-Band accelerator’s short pulse. With this, on-site neutron Time of Flight measurement can be used to identify the nuclear debris isotope contents through Neutron Resonance Transmission Analysis (NRTA). Some preliminary NRTA experiments have been done with Tungsten sample as dummy nuclear debris material, which isotopes Tungsten-186 has close energy absorption value with Uranium-238 (15 eV). The results obtained shows that this system can detect energy absorption in the resonance neutron area within 1-100 eV. It can also detect multiple elements in a material at once with the experiment using a combined sample of Indium, Tantalum, and silver makes it feasible to identify debris containing mixed material. This compact neutron Time of Flight measurement system is a great complementary for dual energy X-Ray Computed Tomography (CT) method that can identify atomic number quantitatively but with 1-mm spatial resolution and high error bar. The combination of these two measurement methods will able to perform on-site nuclear debris screening at Fukushima Daiichi reactor core area, providing the data for nuclear debris activity mapping.

Keywords: neutron source, neutron resonance, nuclear debris, time of flight

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Authors: Dmitry V. Fomichev, Vladimir I. Solonin

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In this paper, experimental and numerical study of hydrodynamic characteristics of the air coolant flow in the test wire-wrapped assembly is presented. The test assembly has 37 rods, which are similar to the real fuel pins of the BREST-OD-300 fuel assemblies geometrically. Air open loop test facility installed at the “Nuclear Power Plants and Installations” department of BMSTU was used to obtain the experimental data. The obtaining altitudinal distribution of static pressure in the near-wall test assembly as well as velocity and temperature distribution of coolant flow in the test sections can give us some new knowledge about the mechanism of formation of the turbulence flow structure in the wire wrapped fuel assemblies. Numerical simulations of the turbulence flow has been accomplished using ANSYS Fluent 14.5. Different non-local turbulence models have been considered, such as standard and RNG k-e models and k-w SST model. Results of numerical simulations of the flow based on the considered turbulence models give the best agreement with the experimental data and help us to carry out strong analysis of flow characteristics.

Keywords: wire-spaces fuel assembly, turbulent flow structure, computation fluid dynamics

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Authors: Hossam Shalabi, George Hadjisophocleous

Abstract:

The Canadian Nuclear Power Plants (NPPs) use some portions of NUREG/CR-6850 in carrying out Fire Probabilistic Risk Assessment (PRA). An assessment for the applicability of NUREG/CR-6850 to CANDU reactors was performed and a CANDU Fire PRA was introduced. There are 19 operating CANDU reactors in Canada at five sites (Bruce A, Bruce B, Darlington, Pickering and Point Lepreau). A fire load density survey was done for all Fire Safe Shutdown Analysis (FSSA) fire zones in all CANDU sites in Canada. National Fire Protection Association (NFPA) Standard 557 proposes that a fire load survey must be conducted by either the weighing method or the inventory method or a combination of both. The combination method results in the most accurate values for fire loads. An updated CANDU Fire PRA model is demonstrated in this paper that includes the fuel survey in all Canadian CANDU stations. A qualitative screening step for the CANDU fire PRA is illustrated in this paper to include any fire events that can damage any part of the emergency power supply in addition to FSSA cables.

Keywords: fire safety, CANDU, nuclear, fuel densities, FDS, qualitative analysis, fire probabilistic risk assessment

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Authors: Reza Bahboodian

Abstract:

One of the most important factors in the use of construction machinery is the fuel consumption cost of this equipment. The use of diesel engines in off-road vehicles is an important source of nitrogen oxides and particulate matter. Emissions of nitrogen oxides and particulate matter 10 in off-road vehicles (construction and mining) may be high. Due to the high cost of fuel, it is necessary to minimize fuel consumption. Factors affecting the fuel consumption of these cars are very diverse. Climate changes such as changes in pressure, temperature, humidity, fuel type selection, type of gearbox used in the car are effective in fuel consumption and pollution, and engine efficiency. In this paper, methods for reducing fuel consumption and pollutants by considering valid European and European standards are examined based on new methods such as hybridization, optimal gear change, adding hydrogen to diesel fuel, determining optimal working fluids, and using oxidation catalysts.

Keywords: improve fuel consumption, construction machinery, pollutant reduction, determining the optimal working cycle

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Authors: Julius Ilawe Osayi, Peter Osifo

Abstract:

Solid waste disposal, such as used tires is a global challenge as well as energy crisis due to rising energy demand amidst price uncertainty and depleting fossil fuel reserves. Therefore, the effectiveness of pyrolysis as a disposal method that can transform used tires into liquid fuel and other end-products has made the process attractive to researchers. Although used tires have been converted to liquid fuel using pyrolysis, there is the need to improve on the liquid fuel properties. Hence, this paper reports the investigation of zeolite NaY synthesized from kaolin, a locally abundant soil material in the Benin metropolis as a suitable catalyst and its effect on the properties of pyrolytic oil produced from used tires. The pyrolysis process was conducted for a range of 1 to 10 wt.% of catalyst concentration to used tire at a temperature of 600 oC, a heating rate of 15oC/min and particle size of 6mm. Although no significant increase in pyrolytic oil yield was observed compared to the previously investigated non-catalytic pyrolysis of a used tire. However, the Fourier transform infrared (FTIR), Nuclear Magnetic Resonance (NMR); and Gas chromatography-mass spectrometry (GC-MS) characterization results revealed the pyrolytic oil to possess an improved physicochemical and fuel properties alongside valuable industrial chemical species. This confirms the possibility of transforming kaolin into a catalyst suitable for improved fuel properties of the liquid fraction obtainable from thermal cracking of hydrocarbon materials.

Keywords: catalytic pyrolysis, fossil fuel, kaolin, pyrolytic oil, used tyres, Zeolite NaY

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Authors: Changyeop Lee, Sewon Kim

Abstract:

Reburning is a useful technology in reducing nitric oxide through injection of a secondary hydrocarbon fuel. In this paper, an experimental study has been conducted to evaluate the effect of fuel lean reburning on NOx/CO reduction in LNG flame. Experiments were performed in flames stabilized by a co-flow swirl burner, which was mounted at the bottom of the furnace. Tests were conducted using LNG gas as the reburn fuel as well as the main fuel. The effects of reburn fuel fraction and injection manner of the reburn fuel were studied when the fuel lean reburning system was applied. The paper reports data on flue gas emissions and temperature distribution in the furnace for a wide range of experimental conditions. At steady state, temperature distribution and emission formation in the furnace have been measured and compared. This paper makes clear that in order to decrease both NOx and CO concentrations in the exhaust when the pulsated fuel lean reburning system was adapted, it is important that the control of some factors such as frequency and duty ratio. Also it shows the fuel lean reburning is also effective method to reduce NOx as much as reburning.

Keywords: fuel lean reburn, NOx, CO, LNG flame

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

Abstract:

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

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

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