Search results for: spent nuclear fuel cladding.

Authors: P. C. Chiu, Jehnming Lin

Abstract:

In this study, a 50-W CO₂ laser was used for the clad of 304L powders on the stainless steel substrate with a temperature sensor and image monitoring system. The laser power and cladding speed and focal position were modified to achieve the requirement of the workpiece flatness and mechanical properties. The numerical calculation is based on ANSYS to analyze the temperature change of the moving heat source at different surface positions when coating the workpiece, and the effect of the process parameters on the bath size was discussed. The temperature of stainless steel powder in the nozzle outlet reacting with the laser was simulated as a process parameter. In the experiment, the difference of the thermal conductivity in three-dimensional space is compared with single-layer cladding and multi-layer cladding. The heat dissipation pattern of the single-layer cladding is the steel plate and the multi-layer coating is the workpiece itself. The relationship between the multi-clad temperature and the profile was analyzed by the temperature signal from an IR pyrometer.

Keywords: laser cladding, temperature, profile, microstructure

Procedia PDF Downloads 221

Authors: Chin-Hsien Yeh, Shao-Wen Chen, Wen-Shu Huang, Chun-Fu Huang, Jong-Rong Wang, Jung-Hua Yang, Yuh-Ming Ferng, Chunkuan Shih

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In this research, a Fukushima-like conditions is simulated with TRACE and RELAP5. Fukushima Daiichi Nuclear Power Plant (NPP) occurred the disaster which caused by the earthquake and tsunami. This disaster caused extended loss of all AC power (ELAP). Hence, loss of ultimate heat sink (LUHS) happened finally. In order to handle Fukushima-like conditions, Taiwan Atomic Energy Council (AEC) commanded that Taiwan Power Company should propose strategies to ensure the nuclear power plant safety. One of the diverse and flexible coping strategies (FLEX) is a different water injection strategy. It can execute core injection at 20 Kg/cm2 without depressurization. In this study, TRACE and RELAP5 were used to simulate Maanshan nuclear power plant, which is a three loops PWR in Taiwan, under Fukushima-like conditions and make sure the success criteria of FLEX. Reducing core cooling ability is due to failure of emergency core cooling system (ECCS) in extended loss of all AC power situation. The core water level continues to decline because of the seal leakage, and then FLEX is used to save the core water level and make fuel rods covered by water. The result shows that this mitigation strategy can cool the reactor pressure vessel (RPV) as soon as possible under Fukushima-like conditions, and keep the core water level higher than Top of Active Fuel (TAF). The FLEX can ensure the peak cladding temperature (PCT) below than the criteria 1088.7 K. Finally, the FLEX can provide protection for nuclear power plant and make plant safety.

Keywords: TRACE, RELAP5/MOD3.3, ELAP, FLEX

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Authors: Heather Wyman-Pain, Yuankai Bian, Furong Li

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The decreasing use of fossil fuel power stations has a negative effect on the stability of the electricity systems in many countries. Nuclear power stations have traditionally provided minimal ancillary services to support the system but this must change in the future as they replace fossil fuel generators. This paper explains the development of the four most popular reactor types still in regular operation across the world which have formed the basis for most reactor development since their commercialisation in the 1950s. The use of nuclear power in four countries with varying levels of capacity provided by nuclear generators is investigated, using the primary frequency response provided by generators as a measure for the electricity networks stability, to assess the need for nuclear generators to provide additional support as their share of the generation capacity increases.

Keywords: frequency control, nuclear power generation, power system stability, system inertia

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Authors: Amir Raeisdana, Davood Sohrabi, Mojtaba Nohekhan, Ameneh Kargarian, Maryam Ghapanvari, Alireza Aslezaeem

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Improvement of zirconium properties by chromium coating and nitrogen implantation is ideal to protect the nuclear fuel rods against corrosion and secondary hydrogenation. Metallic chromium (Cr) has attracted attention as a potential coating material on zirconium alloys, to limit external cladding corrosion. In this research, high energy plasma focus device was used to coat the chromium and implant the nitrogen ions in the zirconium substrate. This device emits high-energy nitrogen ions of 10 keV-1 MeV and with a flux of 10^16 ions/cm^2 in each shot toward the target so it is attractive for implantation on the substrate materials at the room temperature. Six zirconium samples in 2cm×2cm dimensions with 1mm thickness were located at a distance of 20cm from the place where the pinch is formed. The experiments are carried out in 0.5 mbar of the nitrogen gas pressure and 15 kV of the charging voltage. Pure Cr disc was installed on the anode head for sputtering of the chromium and deposition on zirconium substrate. When the pinch plasma column decays due to various instabilities, intense and high-energy N2 ions are accelerated towards the zirconium substrate also sputtered Cr is deposited on the zirconium substrate. XRD and XRF analysis were used to study the structural properties of the samples. XRF analysis indicates 77.1% of Zr and 11.1% of Cr in the surface of the sample. XRD spectra shows the formation of ZrN, CrN and CrZr composites after nitrogen implantation and chromium coating. XRD spectra shows the chromium peak height equal to 152.80 a.u. for the major sample (θ=0֯) and 92.99 a.u. for the minor sample (θ=6֯), so implantation and coating along the main axis of the device is significantly more than other directions.

Keywords: ZrN and CrN and CrZr composites, angular distribution for Cr deposition rate, zirconium corrosion resistance, nuclear fuel rods, plasma focus device

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Authors: Taposh Roy, Anna Paradowska, Ralph Abrahams, Quan Lai, Michael Law, Peter Mutton, Mehdi Soodi, Wenyi Yan

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In this investigation, a laser cladding process with a powder feeding was used to deposit stainless steel 410L (high strength, excellent resistance to abrasion and corrosion, and great laser compatibility) onto railhead (higher strength, heat treated hypereutectoid rail grade manufactured in accordance with the requirements of European standard EN 13674 Part 1 for R400HT grade), to investigate the development and controllability of process-induced residual stress in the cladding, heat-affected zone (HAZ) and substrate and to analyse their correlation with hardness profile during two different laser cladding directions (across and along the track). Residual stresses were analysed by neutron diffraction at OPAL reactor, ANSTO. Neutron diffraction was carried out on the samples in longitudinal (parallel to the rail), transverse (perpendicular to the rail) and normal (through thickness) directions with high spatial resolution through the thickness. Due to the thick rail and thin cladding, 4 mm thick reference samples were prepared from every specimen by Electric Discharge Machining (EDM). Metallography across the laser claded sample revealed four distinct zones: The clad zone, the dilution zone, HAZ and the substrate. Compressive residual stresses were found in the clad zone and tensile residual stress in the dilution zone and HAZ. Laser cladding in longitudinally cladding induced higher tensile stress in the HAZ, whereas transversely cladding rail showed lower tensile behavior.

Keywords: laser cladding, residual stress, neutron diffraction, HAZ

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Authors: Sanghoon Bae, Hanju Cha

Abstract:

Failed fuel detector (FFD) in research reactor is a very crucial instrument to detect the anomaly from failed fuels in the early stage around primary cooling system (PCS) outlet prior to the decay tank. FFD is considered as a mandatory sensor to ensure the integrity of fuel assemblies and mitigate the consequence from a failed fuel accident. For the effective function of FFD, the location of them should be determined by contemplating the effect from coolant flow around two outlets. For this, the analysis on computational flow dynamics (CFD) should be first performed how the coolant outlet flow including radioactive materials from failed fuels are mixed and discharged through the outlet plenum within certain seconds. The analysis result shows that the outlet flow is well mixed regardless of the position of failed fuel and ultimately illustrates the effect of detector location.

Keywords: computational flow dynamics (CFD), failed fuel detector (FFD), fresh fuel assembly (FFA), spent fuel assembly (SFA)

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Authors: Yury S. Shpanskiy, Boris V. Kuteev

Abstract:

Development of a fusion-fission hybrid facility based on superconducting conventional tokamak DEMO-FNS runs in Russia since 2013. The main design goal is to reach the technical feasibility and outline prospects of industrial hybrid technologies providing the production of neutrons, fuel nuclides, tritium, high-temperature heat, electricity and subcritical transmutation in Fusion-Fission Hybrid Systems. The facility should operate in a steady-state mode at the fusion power of 40 MW and fission reactions of 400 MW. Major tokamak parameters are the following: major radius R=3.2 m, minor radius a=1.0 m, elongation 2.1, triangularity 0.5. The design provides the neutron wall loading of ~0.2 MW/m², the lifetime neutron fluence of ~2 MWa/m², with the surface area of the active cores and tritium breeding blanket ~100 m². Core plasma modelling showed that the neutron yield ~10¹⁹ n/s is maximal if the tritium/deuterium density ratio is 1.5-2.3. The design of the electromagnetic system (EMS) defined its basic parameters, accounting for the coils strength and stability, and identified the most problematic nodes in the toroidal field coils and the central solenoid. The EMS generates toroidal, poloidal and correcting magnetic fields necessary for the plasma shaping and confinement inside the vacuum vessel. EMC consists of eighteen superconducting toroidal field coils, eight poloidal field coils, five sections of a central solenoid, correction coils, in-vessel coils for vertical plasma control. Supporting structures, the thermal shield, and the cryostat maintain its operation. EMS operates with the pulse duration of up to 5000 hours at the plasma current up to 5 MA. The vacuum vessel (VV) is an all-welded two-layer toroidal shell placed inside the EMS. The free space between the vessel shells is filled with water and boron steel plates, which form the neutron protection of the EMS. The VV-volume is 265 m³, its mass with manifolds is 1800 tons. The nuclear blanket of DEMO-FNS facility was designed to provide functions of minor actinides transmutation, tritium production and enrichment of spent nuclear fuel. The vertical overloading of the subcritical active cores with MA was chosen as prospective. Analysis of the device neutronics and the hybrid blanket thermal-hydraulic characteristics has been performed for the system with functions covering transmutation of minor actinides, production of tritium and enrichment of spent nuclear fuel. A study of FNS facilities role in the Russian closed nuclear fuel cycle was performed. It showed that during ~100 years of operation three FNS facilities with fission power of 3 GW controlled by fusion neutron source with power of 40 MW can burn 98 tons of minor actinides and 198 tons of Pu-239 can be produced for startup loading of 20 fast reactors. Instead of Pu-239, up to 25 kg of tritium per year may be produced for startup of fusion reactors using blocks with lithium orthosilicate instead of fissile breeder blankets.

Keywords: fusion-fission hybrid system, conventional tokamak, superconducting electromagnetic system, two-layer vacuum vessel, subcritical active cores, nuclear fuel cycle

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Authors: Hamed Djalal

Abstract:

The aim of this paper is to perform a thermal-hydraulic analysis of the IAEA 10 MW benchmark reactor solving analytically and numerically, by mean of the finite volume method, respectively the steady state and transient forced convection in rectangular narrow channel between two parallel MTR-type fuel plates, imposed under a cosine shape heat flux. A comparison between both solutions is presented to determine the minimal coolant velocity which can ensure a safe reactor core cooling, where the cladding temperature should not reach a specific safety limit 90 °C. For this purpose, a 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 inlet coolant velocity. Finally, a good agreement is noticed between the both analytical and numerical solutions, where the obtained results are displayed graphically.

Keywords: forced convection, pressure drop, thermal hydraulic analysis, vertical heated rectangular channel

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Authors: Abid Hossain Khan, Syam Hasan, M. A. R. Sarkar

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Fukushima disaster is one of the most publicly exposed accidents in a nuclear facility which has changed the outlook of people towards nuclear power. Some have used it as an example to establish nuclear energy as an unsafe source, while others have tried to find the real reasons behind this accident. Many papers have tried to shed light on the possible causes, some of which are purely based on assumptions while others rely on rigorous data analysis. To our best knowledge, none of the works can say with absolute certainty that there is a single prominent reason that has paved the way to this unexpected incident. This paper attempts to compile all the apparent reasons behind Fukushima disaster and tries to analyze and identify the most likely one.

Keywords: fuel meltdown, Fukushima disaster, Manmade calamity, nuclear facility, tsunami

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Authors: Meena Marafi, Mohan S. Rana

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Catalysts play an important role in producing valuable fuel products in petroleum refining; but, due to feedstock’s impurities catalyst gets deactivated with carbon and metal deposition. The disposal of spent catalyst falls under the category of hazardous industrial waste that requires strict agreement with environmental regulations. The spent hydroprocessing catalyst contains Mo, V and Ni at high concentrations that have been found to be economically significant for recovery. Metal recovery process includes deoiling, decoking, grinding, dissolving and treatment with complexing leaching agent such as ethylene diamine tetra acetic acid (EDTA). The process conditions have been optimized as a function of time, temperature and EDTA concentration in presence of ultrasonic agitation. The results indicated that optimum condition established through this approach could recover 97%, 94% and 95% of the extracted Mo, V and Ni, respectively, while 95% EDTA was recovered after acid treatment.

Keywords: atmospheric residue desulfurization (ARDS), deactivation, hydrotreating, spent catalyst

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Authors: R. S. Rajpurohit, N. C. Santhi Srinivas, Vakil Singh

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Asymmetric stress cycling leads to accumulation of plastic strain which is called as ratcheting strain. The problem is generally associated with nuclear fuel cladding materials used in nuclear power plants and pressurized pipelines. In the present investigation, asymmetric stress controlled fatigue tests were conducted with three different parameters namely, mean stress, stress amplitude and stress rate (keeping two parameters constant and varying third parameter) to see the plastic strain accumulation and its effect on fatigue life and deformation behavior of Zircaloy-2 at 400°C. The tests were conducted with variable mean stress (45-70 MPa), stress amplitude (95-120 MPa) and stress rate (30-750 MPa/s) and tested specimens were characterized using transmission and scanning electron microscopy. The experimental results show that with the increase in mean stress and stress amplitude, the ratcheting strain accumulation increases with reduction in fatigue life. However, increase in stress rate leads to improvement in fatigue life of the material due to small ratcheting strain accumulation. Fractographs showed a decrease in area fraction of fatigue failed region.

Keywords: asymmetric cyclic loading, ratcheting fatigue, mean stress, stress amplitude, stress rate, plastic strain

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Authors: Rafael Oliveira Santos, Luciano Pessanha Moreira, Marcelo Costa Cardoso

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Spacer grid assembly supporting the nuclear fuel rods is an important concern in the design of structural components of a Pressurized Water Reactor (PWR). The spacer grid is composed by springs and dimples which are formed from a strip sheet by means of blanking and stamping processes. In this paper, the blanking process and tooling parameters are evaluated by means of a 2D plane-strain finite element model in order to evaluate the punch load and quality of the sheared edges of Inconel 718 strips used for nuclear spacer grids. A 3D finite element model is also proposed to predict the tooling loads resulting from the stamping process of a preformed Inconel 718 strip and to analyse the residual stress effects upon the spring and dimple design geometries of a nuclear spacer grid.

Keywords: blanking process, damage model, finite element modelling, inconel 718, spacer grids, stamping process

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

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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

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Authors: S. A. H. Feghhi, Z. Gholamzadeh, Z. Alipoor, C. Tenreiro

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Currently, thorium fuel has been especially noticed because of its proliferation resistance than long half-life alpha emitter minor actinides, breeding capability in fast and thermal neutron flux and mono-isotopic naturally abundant. In recent years, efficiency of minor actinide burning up in PWRs has been investigated. Hence, a minor actinide-contained thorium based fuel matrix can confront both proliferation resistance and nuclear waste depletion aims. In the present work, minor actinide depletion rate in a CANDU-type nuclear core modeled using MCNP code has been investigated. The obtained effects of minor actinide load as mixture of thorium fuel matrix on the core neutronics has been studiedwith comparingpresence and non-presence of minor actinide component in the fuel matrix.Depletion rate of minor actinides in the MA-contained fuel has been calculated using different power loads.According to the obtained computational data, minor actinide loading in the modeled core results in more negative reactivity coefficients. The MA-contained fuel achieves less radial peaking factor in the modeled core. The obtained computational results showed 140 kg of 464 kg initial load of minor actinide has been depleted in during a 6-year burn up in 10 MW power.

Keywords: minor actinide burning, CANDU-type reactor, MCNPX code, neutronic parameters

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Authors: Sami I. Jafar, Sami A. Ajeel, Zaman A. Abdulwahab

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The turbine blades used in electric power plants are made of low alloy steel type 52. These blades will be subjected to fatigue and also at other times to fatigue corrosion with aging time. Due to their continuous exposure to cyclic rotational stresses in corrosive steam environments, The current research aims to deal with this problem using the laser cladding method for low alloy steel type 52, which works to re-compose the metallurgical structure and improve the mechanical properties by strengthening the resulting structure, which leads to an increase in fatigue and wears resistance, therefore, an increase in the life of these blades is observed.

Keywords: fatigue, fatigue corrosion, turbine blades, laser cladding

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Authors: Haragobinda Srichandan, Ashish Pathak, Dong Jin Kim, Seoung-Won Lee

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The present investigation deals with bioleaching of spent petroleum catalyst using At. ferrooxidans and At. thiooxidans. The spent catalyst used in the present study was pretreated with acetone to remove the oily hydrocarbons. FESEM and XPS analysis indicated the presence of metals in sulfide and oxide forms in spent catalyst. Both At. ferrooxidans and At. thiooxidans were found to be highly effective in producing the acid. Bioleaching with At. ferrooxidans and At. thiooxidans led to higher recovery of metals compare to control. During bioleaching similar recoveries of metals were obtained using At. ferrooxidans and At. thiooxidans. This might be due to the presence of metals as soluble oxides and sulphides in the spent catalyst. At the end of bioleaching, about 87-90% Ni, 34% Al, 65-73% Mo and 92-97% V were leached using above bacteria. It is elucidated that bioleaching with At. thiooxidans is comparatively more advantageous due to lower cost of sulphur.

Keywords: At. ferrooxidans, bioleaching, metal recovery, spent catalyst

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Authors: Jamal Al-Zain, O. El Hajjaji, T. El Bardouni

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A (Th-U) O2 fuel pin benchmark made up of 25 w/o U and 75 w/o Th was used. In order to analyze the depletion and inventory of the fuel for the pressurized water reactor pin-cell model. The new version VIII.0 of the ENDF/B nuclear data library was used to create a data set in ACE format at various temperatures and process the data using the MAKXSF6.2 and NJOY2016 programs to process the data at the various temperatures in order to conduct this study and analyze cross-section data. The infinite multiplication factor, the concentrations and activities of the main fission products, the actinide radionuclides accumulated in the pin cell, and the total radioactivity were all estimated and compared in this study using the Monte Carlo N-Particle 6 (MCNP6.2) and DRAGON5 programs. Additionally, the behavior of the Pressurized Water Reactor (PWR) thorium pin cell that is dependent on burn-up (BU) was validated and compared with the reference data obtained using the Massachusetts Institute of Technology (MIT-MOCUP), Idaho National Engineering and Environmental Laboratory (INEEL-MOCUP), and CASMO-4 codes. The results of this study indicate that all of the codes examined have good agreements.

Keywords: PWR thorium pin cell, ENDF/B-VIII.0, MAKXSF6.2, NJOY2016, MCNP6.2, DRAGON5, fuel burn-up.

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Authors: Harish Aryal

Abstract:

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

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

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Authors: Yuan-Ching Lin, Sin-Yu Chen, Pei-Yu Wu

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This study reports the improvement in the wear performance of A6061 aluminum alloy clad with mixed powders of titanium carbide (TiC), copper (Cu) and aluminum (Al) using the gas tungsten arc welding (GTAW) method. The wear performance of the A6061 clad layers was evaluated by performing pin-on-disc mode wear test. Experimental results clearly indicate an enhancement in the hardness of the clad layer by about two times that of the A6061 substrate without cladding. Wear test demonstrated a significant improvement in the wear performance of the clad layer when compared with the A6061 substrate without cladding. Moreover, the interface between the clad layer and the A6061 substrate exhibited superior metallurgical bonding. Due to this bonding, the clad layer did not spall during the wear test; as such, massive wear loss was prevented. Additionally, massive oxidized particulate debris was generated on the worn surface during the wear test; this resulted in three-body abrasive wear and reduced the wear behavior of the clad surface.

Keywords: GTAW、A6061 aluminum alloy, 、surface modification, tribological study, TiC powder cladding

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Authors: K. A. Sudjatmi, Endiah Puji Hastuti, Surip Widodo, Reinaldy Nazar

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Any pretensions to stop the production of TRIGA fuel elements by TRIGA reactor fuel elements manufacturer should be anticipated by the operating agency of TRIGA reactor to replace the cylinder type fuel element with plate type fuel element, that available on the market. This away was performed the calculation on U3Si2Al fuel with uranium enrichment of 19.75% and a load level of 2.96 gU/cm3. Maximum power that can be operated on free convection cooling mode at the BANDUNG TRIGA reactor fuel plate was 600 kW. This study has been conducted thermalhydraulic characteristic calculation model of the reactor core power 2MW. BANDUNG TRIGA reactor core fueled plate type is composed of 16 fuel elements, 4 control elements and one irradiation facility which is located right in the middle of the core. The reactor core is cooled using a pump which is already available with flow rate 900 gpm. Analysis on forced convection cooling mode with flow from the top down from 10%, 20%, 30% and so on up to a 100% rate of coolant flow. performed using the COOLOD-N2 code. The calculations result showed that the 2 MW power with inlet coolant temperature at 37 °C and cooling rate percentage of 50%, then the coolant temperature, maximum cladding and meat respectively 64.96 oC, 124.81 oC, and 125.08 oC, DNBR (departure from nucleate boiling ratio)=1.23 and OFIR (onset of flow instability ratio)=1:00. The results are expected to be used as a reference for determining the power and cooling rate level of the BANDUNG TRIGA reactor core plate types fueled.

Keywords: TRIGA, COOLOD-N2, plate type fuel element, force convection, thermal hydraulic characteristic

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

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In this study of ultimate response guideline (URG), Kuosheng BWR/6 nuclear power plant (NPP) TRACE model was established. The reactor depressurization, low pressure water injection, and containment venting are the main actions of URG. This research focuses to evaluate the efficiency of URG under Fukushima-like conditions. Additionally, the sensitivity study of URG was also performed in this research. The analysis results of TRACE present that URG can keep the peak cladding temperature (PCT) below 1088.7 K (the failure criteria) under Fukushima-like conditions. It implied that Kuosheng NPP was at the safe situation.

Keywords: BWR, TRACE, safety analysis, ultimate response guideline (URG)

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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: Dimitrie Marinceu, Alan Murchison

Abstract:

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: Pinar Temocin

Abstract:

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: 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: Alain Kusmoko, Druce Dunne, Huijun Li

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Stellite 6 was deposited by laser cladding on a chromium bearing substrate (P91) with energy inputs of 1 kW (P91-1) and 1.8 kW (P91-1.8). The chemical compositions and microstructures of these coatings were characterized by atomic absorption spectroscopy, optical microscopy and scanning electron microscopy. The microhardness of the coatings was measured and the wear mechanism of the coatings was assessed using a pin-on-plate (reciprocating) wear testing machine. The results showed less cracking and pore development for Stellite 6 coatings applied to the P91 steel substrate with the lower heat input (P91-1). Further, the Stellite coating for P91-1 was significantly harder than that obtained for P91-1.8. The wear test results indicated that the weight loss for P91-1 was much lower than for P91-1.8. It is concluded that the lower hardness of the coating for P91-1.8, together with the softer underlying substrate structure, markedly reduced the wear resistance of the Stellite 6 coating.

Keywords: friction and wear, laser cladding, P91 steel, Stellite 6 coating

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Authors: Asil Zureigat, Ayat Odat

Abstract:

Analyzing the old and bringing in the new is an ever ongoing process in driving innovations in architecture. This paper looks at the excessive use of stone in apartment buildings in Amman and speculates on the existing possibilities of changing the cladding material. By looking at architectural exceptions present in Amman the paper seeks to make the exception, the rule by adding new materials to the architectural library of Amman and in turn, project a series of possible new identities to the existing stone scape. Through distributing a survey, conducting a photographic study on exceptional buildings and shedding light on the historical narrative of stone, the paper highlights the ways in which new finishing materials such as plaster, paint and stone variations could be introduced in an attempt to project a new architectural identity to Amman.

Keywords: architectural city identity, cladding materials, façade architecture, image of the city

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Authors: Si-Geun Choi, Hoon-Jae Park, Jung-Woo Cho, Jin-Ho Lim, Jin-Young Park, Joo-Young Oh, Jae-Il Jeong Seock-Sam Kim, Young Tae Cho, Chan Gyu Kim, Jong-Hyoung Kim

Abstract:

The tilting pad bearing is a kind of the fluid film bearing and it can contribute to the high speed and the high load performance compared to other bearings including the rolling element bearing. Furthermore, the tilting bearing has many advantages such as high stability at high-speed performance, long life, high damping, high impact resistance and low noise. Therefore, it mostly used in mid to large size turbomachines, despite the high price disadvantage. Recently, manufacture and process employing laser techniques advancing at a fast-growing rate in mechanical industry, the dissimilar metal weld process employing laser techniques is actively studied. Moreover, also, Industry fields try to apply for welding the white metal and the back metal using laser cladding method for high durability. Furthermore, it has followed that laser cladding method has a lot better bond strength, toughness, anti-abrasion and environment-friendly than centrifugal casting method through preceding research. Therefore, the laser cladding method has a lot better quality, cost reduction, eco-friendliness and permanence of technology than the centrifugal casting method or the gravity casting method. In this study, we compare the mechanical properties of different bearing materials by evaluating the behavior of laser cladding layer with various materials (i.e. SS400, SCM440, S20C) under the same parameters. Furthermore, we analyze the porosity of various tilting pad bearing materials which white metal treated on samples. SEM, EDS analysis and hardness tests of three materials are shown to understand the mechanical properties and tribological behavior. W/D ratio, surface roughness results with various materials are performed in this study.

Keywords: laser cladding, tilting pad bearing, white metal, mechanical properties

Procedia PDF Downloads 377

Authors: Alireza Vakili, Shahab Ehtesham, Mohsen Danesh Mesgaran

Abstract:

The objective of this study was use different levels of spent mushroom substrate as a suitable substitute for wheat straw in the ration of male lambs. In this study 20 male lambs with the age of 90 days and initial average weight of 33± 1.7 kg were used. The animals were divided separately into single boxes with four treatments (control treatment, spent mushroom substrate 15%, spent mushroom substrate 25% and spent mushroom substrate 35%) and five replications. The experiment period was 114 days being 14 days adaptation and 90 days for breeding. On the days 36 and 94, blood samples were taken from the jugular vein. In order to carry out the trial, 20 male lambs received the four experimental diets in completely randomized design. The statistical analyses were carried out by using the GLM procedure of SAS 9.1. Means among treatments were compared by Tukey test. The results of the study showed that there was no significant differences between the serum biochemical and hematological contents of the lambs in the four treatments (p>0.05). It was concluded that spent mushroom substrate consumption has no harmful effect on the blood parameters of Kurdish male lambs.

Keywords: alternative food, nutrition, sheep performance, spent mushroom substrate

Procedia PDF Downloads 351

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

Abstract:

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

Procedia PDF Downloads 155