Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 7607

Search results for: pressurized water reactor (PWR)

7607 Structural Integrity Analysis of Baffle Former Assembly in Pressurized Water Reactors Considering Irradiation Aging

Authors: Jong-Sung Kim, Myung-Jo Jhung


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

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

Procedia PDF Downloads 294
7606 Using TRACE and SNAP Codes to Establish the Model of Maanshan PWR for SBO Accident

Authors: B. R. Shen, J. R. Wang, J. H. Yang, S. W. Chen, C. Shih, Y. Chiang, Y. F. Chang, Y. H. Huang


In this research, TRACE code with the interface code-SNAP was used to simulate and analyze the SBO (station blackout) accident which occurred in Maanshan PWR (pressurized water reactor) nuclear power plant (NPP). There are four main steps in this research. First, the SBO accident data of Maanshan NPP were collected. Second, the TRACE/SNAP model of Maanshan NPP was established by using these data. Third, this TRACE/SNAP model was used to perform the simulation and analysis of SBO accident. Finally, the simulation and analysis of SBO with mitigation equipments was performed. The analysis results of TRACE are consistent with the data of Maanshan NPP. The mitigation equipments of Maanshan can maintain the safety of Maanshan in the SBO according to the TRACE predictions.

Keywords: pressurized water reactor (PWR), TRACE, station blackout (SBO), Maanshan

Procedia PDF Downloads 127
7605 Neural Network Supervisory Proportional-Integral-Derivative Control of the Pressurized Water Reactor Core Power Load Following Operation

Authors: Derjew Ayele Ejigu, Houde Song, Xiaojing Liu


This work presents the particle swarm optimization trained neural network (PSO-NN) supervisory proportional integral derivative (PID) control method to monitor the pressurized water reactor (PWR) core power for safe operation. The proposed control approach is implemented on the transfer function of the PWR core, which is computed from the state-space model. The PWR core state-space model is designed from the neutronics, thermal-hydraulics, and reactivity models using perturbation around the equilibrium value. The proposed control approach computes the control rod speed to maneuver the core power to track the reference in a closed-loop scheme. The particle swarm optimization (PSO) algorithm is used to train the neural network (NN) and to tune the PID simultaneously. The controller performance is examined using integral absolute error, integral time absolute error, integral square error, and integral time square error functions, and the stability of the system is analyzed by using the Bode diagram. The simulation results indicated that the controller shows satisfactory performance to control and track the load power effectively and smoothly as compared to the PSO-PID control technique. This study will give benefit to design a supervisory controller for nuclear engineering research fields for control application.

Keywords: machine learning, neural network, pressurized water reactor, supervisory controller

Procedia PDF Downloads 77
7604 Probabilistic Fracture Evaluation of Reactor Pressure Vessel Subjected to Pressurized Thermal Shock

Authors: Jianguo Chen, Fenggang Zang, Yu Yang, Liangang Zheng


Reactor Pressure Vessel (RPV) is an important security barrier in nuclear power plant. Crack like defects may be produced on RPV during the whole operation lifetime due to the harsh operation condition and irradiation embrittlement. During the severe loss of coolant accident, thermal shock happened as the injection of emergency cooling water into RPV, which results in re-pressurization of the vessel and very high tension stress on the vessel wall, this event called Pressurized Thermal Shock (PTS). Crack on the vessel wall may propagate even penetrate the vessel, so the safety of the RPV would undergo great challenge. Many assumptions in structure integrity evaluation make the result of deterministic fracture mechanics very conservative, which affect the operation lifetime of the plant. Actually, many parameters in the evaluation process, such as fracture toughness and nil-ductility transition temperature, have statistical distribution characteristics. So it is necessary to assess the structural integrity of RPV subjected to PTS event by means of Probabilistic Fracture Mechanics (PFM). Structure integrity evaluation methods of RPV subjected to PTS event are summarized firstly, then evaluation method based on probabilistic fracture mechanics are presented by considering the probabilistic characteristics of material and structure parameters. A comprehensive analysis example is carried out at last. The results show that the probability of crack penetrates through wall increases gradually with the growth of fast neutron irradiation flux. The results give advice for reactor life extension.

Keywords: fracture toughness, integrity evaluation, pressurized thermal shock, probabilistic fracture mechanics, reactor pressure vessel

Procedia PDF Downloads 180
7603 Environmental Effect on Corrosion Fatigue Behaviors of Steam Generator Forging in Simulated Pressurized Water Reactor Environment

Authors: Yakui Bai, Chen Sun, Ke Wang


An experimental investigation of environmental effect on fatigue behavior in SA508 Gr.3 Cl.2 Steam Generator Forging CAP1400 nuclear power plant has been carried out. In order to simulate actual loading condition, a range of strain amplitude was applied in different low cycle fatigue (LCF) tests. The current American Society of Mechanical Engineers (ASME) design fatigue code does not take full account of the interactions of environmental, loading, and material's factors. A range of strain amplitude was applied in different low cycle fatigue (LCF) tests at a strain rate of 0.01%s⁻¹. A design fatigue model was constructed by taking environmentally assisted fatigue effects into account, and the corresponding design curves were given for the convenience of engineering applications. The corrosion fatigue experiment was performed in a strain control mode in 320℃ borated and lithiated water environment to evaluate the effects of a mixed environment on fatigue life. Stress corrosion cracking (SCC) in steam generator large forging in primary water of pressurized water reactor was also observed. In addition, it is found that the CF life of SA508 Gr.3 Cl.2 decreases with increasing temperature in the water environment. The relationship between the reciprocal of temperature and the logarithm of fatigue life was found to be linear. Through experiments and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for steam generator forging.

Keywords: failure behavior, low alloy steel, steam generator forging, stress corrosion cracking

Procedia PDF Downloads 61
7602 Hard Sludge Formation and Consolidation in Pressurized Water Reactor Steam Generators: An Experimental Study

Authors: R. Fernandez-Saavedra, M. B. Gomez-Mancebo, D. Gomez-Briceno


The gradual corrosion of PWR (Pressurized Water Reactor) feedwater, condensate and drain systems results in the inevitable liberation of corrosion products, principally metallic oxides, to the secondary circuit. In addition, other contaminants and impurities are introduced into the makeup water, auxiliary feedwater and by condenser leaks. All these compounds circulating in the secondary flow can eventually be transported to steam generators and be transformed into deposits on their surfaces. Deposits that accumulate on the tube sheet are known as sludge piles and when they consolidate and harden become into hard sludge. Hard sludge is especially detrimental because it favors tube deformation or denting at the top of tube sheet and further stress corrosion cracking (SCC). These failures affect the efficiency of nuclear power plants. In a recent work, a model for the formation and consolidation of hard sludge has been formulated, highlighting the influence of aluminum and silicon compounds in the initial formation of hard sludge. In this work, an experimental study has been performed in order to get a deeper understanding of the behavior of Al and Si species in hard sludge formation and consolidation. For this purpose, the key components of hard sludge (magnetite, aluminum and/or silicon sources) have been isothermally autoclaved in representative secondary circuit conditions during one week, and the resulting products have been chemically and structurally characterized by XRF and XRD techniques, respectively.

Keywords: consolidation, hard sludge, secondary circuit, steam generator

Procedia PDF Downloads 127
7601 Thermal Hydraulic Analysis of Sub-Channels of Pressurized Water Reactors with Hexagonal Array: A Numerical Approach

Authors: Md. Asif Ullah, M. A. R. Sarkar


This paper illustrates 2-D and 3-D simulations of sub-channels of a Pressurized Water Reactor (PWR) having hexagonal array of fuel rods. At a steady state, the temperature of outer surface of the cladding of fuel rod is kept about 1200°C. The temperature of this isothermal surface is taken as boundary condition for simulation. Water with temperature of 290°C is given as a coolant inlet to the primary water circuit which is pressurized upto 157 bar. Turbulent flow of pressurized water is used for heat removal. In 2-D model, temperature, velocity, pressure and Nusselt number distributions are simulated in a vertical sectional plane through the sub-channels of a hexagonal fuel rod assembly. Temperature, Nusselt number and Y-component of convective heat flux along a line in this plane near the end of fuel rods are plotted for different Reynold’s number. A comparison between X-component and Y-component of convective heat flux in this vertical plane is analyzed. Hexagonal fuel rod assembly has three types of sub-channels according to geometrical shape whose boundary conditions are different too. In 3-D model, temperature, velocity, pressure, Nusselt number, total heat flux magnitude distributions for all the three sub-channels are studied for a suitable Reynold’s number. A horizontal sectional plane is taken from each of the three sub-channels to study temperature, velocity, pressure, Nusselt number and convective heat flux distribution in it. Greater values of temperature, Nusselt number and Y-component of convective heat flux are found for greater Reynold’s number. X-component of convective heat flux is found to be non-zero near the bottom of fuel rod and zero near the end of fuel rod. This indicates that the convective heat transfer occurs totally along the direction of flow near the outlet. As, length to radius ratio of sub-channels is very high, simulation for a short length of the sub-channels are done for graphical interface advantage. For the simulations, Turbulent Flow (K-Є ) module and Heat Transfer in Fluids (ht) module of COMSOL MULTIPHYSICS 5.0 are used.

Keywords: sub-channels, Reynold’s number, Nusselt number, convective heat transfer

Procedia PDF Downloads 291
7600 Nuclear Power Plant Radioactive Effluent Discharge Management in China

Authors: Jie Yang, Qifu Cheng, Yafang Liu, Zhijie Gu


Controlled emissions of effluent from nuclear power plants are an important means of ensuring environmental safety. In order to fully grasp the actual discharge level of nuclear power plant in China's nuclear power plant in the pressurized water reactor and heavy water reactor, it will use the global average nuclear power plant effluent discharge as a reference to the standard analysis of China's nuclear power plant environmental discharge status. The results show that the average normalized emission of liquid tritium in PWR nuclear power plants in China is slightly higher than the global average value, and the other nuclides emissions are lower than the global average values.

Keywords: radioactive effluent, HWR, PWR, nuclear power plant

Procedia PDF Downloads 163
7599 Real-Time Optimisation and Minimal Energy Use for Water and Environment Efficient Irrigation

Authors: Kanya L. Khatri, Ashfaque A. Memon, Rod J. Smith, Shamas Bilal


The viability and sustainability of crop production is currently threatened by increasing water scarcity. Water scarcity problems can be addressed through improved water productivity and the options usually presumed in this context are efficient water use and conversion of surface irrigation to pressurized systems. By replacing furrow irrigation with drip or centre pivot systems, the water efficiency can be improved by up to 30 to 45%. However, the installation and application of pumps and pipes, and the associated fuels needed for these alternatives increase energy consumption and cause significant greenhouse gas emissions. Hence, a balance between the improvement in water use and the potential increase in energy consumption is required keeping in view adverse impact of increased carbon emissions on the environment. When surface water is used, pressurized systems increase energy consumption substantially, by between 65% to 75%, and produce greenhouse gas emissions around 1.75 times higher than that of gravity based irrigation. With gravity based surface irrigation methods the energy consumption is assumed to be negligible. This study has shown that a novel real-time infiltration model REIP has enabled implementation of real-time optimization and control of surface irrigation and surface irrigation with real-time optimization has potential to bring significant improvements in irrigation performance along with substantial water savings of 2.92 ML/ha which is almost equivalent to that given by pressurized systems. Thus real-time optimization and control offers a modern, environment friendly and water efficient system with close to zero increase in energy consumption and minimal greenhouse gas emissions.

Keywords: pressurised irrigation, carbon emissions, real-time, environmentally-friendly, REIP

Procedia PDF Downloads 425
7598 Numerical Simulation of Solar Reactor for Water Disinfection

Authors: A. Sebti Bouzid, S. Igoud, L. Aoudjit, H. Lebik


Mathematical modeling and numerical simulation have emerged over the past two decades as one of the key tools for design and optimize performances of physical and chemical processes intended to water disinfection. Water photolysis is an efficient and economical technique to reduce bacterial contamination. It exploits the germicidal effect of solar ultraviolet irradiation to inactivate pathogenic microorganisms. The design of photo-reactor operating in continuous disinfection system, required tacking in account the hydrodynamic behavior of water in the reactor. Since the kinetic of disinfection depends on irradiation intensity distribution, coupling the hydrodynamic and solar radiation distribution is of crucial importance. In this work we propose a numerical simulation study for hydrodynamic and solar irradiation distribution in a tubular photo-reactor. We have used the Computational Fluid Dynamic code Fluent under the assumption of three-dimensional incompressible flow in unsteady turbulent regimes. The results of simulation concerned radiation, temperature and velocity fields are discussed and the effect of inclination angle of reactor relative to the horizontal is investigated.

Keywords: solar water disinfection, hydrodynamic modeling, solar irradiation modeling, CFD Fluent

Procedia PDF Downloads 271
7597 Stress Analysis of a Pressurizer in a Pressurized Water Reactor Using Finite Element Method

Authors: Tanvir Hasan, Minhaz Uddin, Anwar Sadat Anik


A pressurizer is a safety-related reactor component that maintains the reactor operating pressure to guarantee safety. Its structure is usually made of high thermal and pressure resistive material. The mechanical structure of these components should be maintained in all working settings, including transient to severe accidents conditions. The goal of this study is to examine the structural integrity and stress of the pressurizer in order to ensure its design integrity towards transient situations. For this, the finite element method (FEM) was used to analyze the mechanical stress on pressurizer components in this research. ANSYS MECHANICAL tool was used to analyze a 3D model of the pressurizer. The material for the body and safety relief nozzle is selected as low alloy steel i.e., SA-508 Gr.3 Cl.2. The model was put into ANSYS WORKBENCH and run under the boundary conditions of (internal Pressure, -17.2 MPa, inside radius, -1348mm, the thickness of the shell, -127mm, and the ratio of the outside radius to an inside radius, - 1.059). The theoretical calculation was done using the formulas and then the results were compared with the simulated results. When stimulated at design conditions, the findings revealed that the pressurizer stress analysis completely fulfilled the ASME standards.

Keywords: pressurizer, stress analysis, finite element method, nuclear reactor

Procedia PDF Downloads 70
7596 Feasibility Study to Enhance the Heat Transfer in a Typical Pressurized Water Reactor by Ribbed Spacer Grids

Authors: A. Ghadbane, M. N. Bouaziz, S. Hanini, B. Baggoura, M. Abbaci


The spacer grids are used to fix the rods bundle in a nuclear reactor core also act as turbulence-enhancing devices to improve the heat transfer from the hot surfaces of the rods to the surrounding coolant stream. Therefore, the investigation of thermal-hydraulic characteristics inside the rod bundles is important for optima design and safety operation of a nuclear reactor power plant. This contribution presents a feasibility study to use the ribbed spacer grids as mixing devices. The present study evaluates the effects of different ribbed spacer grids configurations on flow pattern and heat transfer in the downstream of the mixing devices in a 2 x 2 rod bundle array. This is done by obtaining velocity and pressure fields, turbulent intensity and the heat transfer coefficient using a three-dimensional CFD analysis. Numerical calculations are performed by employing K-ε turbulent model. The computational results obtained are promising and the comparison with standard spacer grids shows a clear difference which required the experimental approach to validate.

Keywords: PWR fuel assembly, spacer grid, mixing vane, swirl flow, turbulent heat transfer, CFD

Procedia PDF Downloads 414
7595 The Main Steamline Break Transient Analysis for Advanced Boiling Water Reactor Using TRACE, PARCS, and SNAP Codes

Authors: H. C. Chang, J. R. Wang, A. L. Ho, S. W. Chen, J. H. Yang, C. Shih, L. C. Wang


To confirm the reactor and containment integrity of the Advanced Boiling Water Reactor (ABWR), we perform the analysis of main steamline break (MSLB) transient by using the TRACE, PARCS, and SNAP codes. The process of the research has four steps. First, the ABWR nuclear power plant (NPP) model is developed by using the above codes. Second, the steady state analysis is performed by using this model. Third, the ABWR model is used to run the analysis of MSLB transient. Fourth, the predictions of TRACE and PARCS are compared with the data of FSAR. The results of TRACE/PARCS and FSAR are similar. According to the TRACE/PARCS results, the reactor and containment integrity of ABWR can be maintained in a safe condition for MSLB.

Keywords: advanced boiling water reactor, TRACE, PARCS, SNAP

Procedia PDF Downloads 139
7594 Neutronic Calculations for Central Test Loop in Heavy Water Research Reactor

Authors: Hadi Shamoradifar, Behzad Teimuri, Parviz Parvaresh, Saeed Mohammadi


One of the experimental facilities of the heavy water research reactor is the central test loop (C.T.L). It is located along the central axial line of the vessel, and therefore will highly affect the neutronic parameters of the reactor, so from the neutronics point of view, C.T.L is the most important facility. It is mainly designed for fuel testing, thought other applications such as radioisotope production and neutron activation, can be imagine for it. All of the simulations were performed by MCNPX2.6. As a first step towards C.T.L analysis, the effect of D2O-filled, H2O-filled, and He-filled C.T.L on the effective multiplication factor (Keff.), have been evaluated. According to results, H2O-filled C.T.L has a higher thermal neutron, while He-filled C.T.L includes more resonance neutrons. In the next step thermal and total axial neutron fluxes, were calculated and used as the comparison parameters. The core without C.T.L (C.T.L replaced by heavy water) is selected as the reference case, and the effect of all other cases is calculated according to that.

Keywords: heavy water reactor, neutronic calculations, central test loop, neutron activation

Procedia PDF Downloads 289
7593 Process Safety Evaluation of a Nuclear Power Plant through Virtual Process Hazard Analysis Using Hazard and Operability Technique

Authors: Elysa V. Largo, Lormaine Anne A. Branzuela, Julie Marisol D. Pagalilauan, Neil C. Concibido, Monet Concepcion M. Detras


The energy demand in the country is increasing; thus, nuclear energy is recently mandated to add to the energy mix. The Philippines has the Bataan Nuclear Power Plant (BNPP), which can be a source of nuclear energy; however, it has not been operated since the completion of its construction. Thus, evaluating the safety of BNPP is vital. This study explored the possible deviations that may occur in the operation of a nuclear power plant with a pressurized water reactor, which is similar to BNPP, through a virtual process hazard analysis (PHA) using the hazard and operability (HAZOP) technique. Temperature, pressure, and flow were used as parameters. A total of 86 causes of various deviations were identified, wherein the primary system and line from reactor coolant pump to reactor vessel are the most critical system and node, respectively. A total of 348 scenarios were determined. The critical events are radioactive leaks due to nuclear meltdown and sump overflow that could lead to multiple worker fatalities, one or more public fatalities, and environmental remediation. There were existing safeguards identified; however, further recommendations were provided to have additional and supplemental barriers to reduce the risk.

Keywords: PSM, PHA, HAZOP, nuclear power plant

Procedia PDF Downloads 12
7592 Integrating Deterministic and Probabilistic Safety Assessment to Decrease Risk & Energy Consumption in a Typical PWR

Authors: Ebrahim Ghanbari, Mohammad Reza Nematollahi


Integrating deterministic and probabilistic safety assessment (IDPSA) is one of the most commonly used issues in the field of safety analysis of power plant accident. It has also been recognized today that the role of human error in creating these accidents is not less than systemic errors, so the human interference and system errors in fault and event sequences are necessary. The integration of these analytical topics will be reflected in the frequency of core damage and also the study of the use of water resources in an accident such as the loss of all electrical power of the plant. In this regard, the SBO accident was simulated for the pressurized water reactor in the deterministic analysis issue, and by analyzing the operator's behavior in controlling the accident, the results of the combination of deterministic and probabilistic assessment were identified. The results showed that the best performance of the plant operator would reduce the risk of an accident by 10%, as well as a decrease of 6.82 liters/second of the water sources of the plant.

Keywords: IDPSA, human error, SBO, risk

Procedia PDF Downloads 50
7591 The Influence of Temperature on Apigenin Extraction from Chamomile (Matricaria recutita) by Superheated Water

Authors: J. Švarc-Gajić, A. Cvetanović


Apigenin is a flavone synthetized by many plants and quite abundant in chamomile (Matricaria recutita) in its free form and in the form of its glucoside and different acylated forms. Many beneficial health effects have been attributed to apigenin, such as chemo-preventive, anxiolytic, anti-inflammatory, antioxidant and antispasmodic. It is reported that free apigenin is much more bioactive in comparison to its bound forms. Subcritical water offers numerous advantages in comparison to conventional extraction techniques, such as good selectivity, low price and safety. Superheated water exhibits high hydrolytical potential which must be carefully balanced when using this solvent for the extraction of bioactive molecules. Moderate hydrolytical potential can be exploited to liberate apigenin from its bound forms, thus increasing biological potential of obtained extracts. The polarity of pressurized water and its hydrolytical potential are highly dependent on the temperature. In this research chamomile ligulate flowers were extracted by pressurized hot water in home-made subcritical water extractor in conditions of convective mass transfer. The influence of the extraction temperature was investigated at 30 bars. Extraction yields of total phenols, total flavonoids and apigenin depending on the operational temperature were calculated based on spectrometric assays. Optimal extraction temperature for maximum yields of total phenols and flavonoids showed to be 160°C, whereas apigenin yield was the highest at 120°C.

Keywords: superheated water, temperature, chamomile, apigenin

Procedia PDF Downloads 428
7590 A Real Time Expert System for Decision Support in Nuclear Power Plants

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


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

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

Procedia PDF Downloads 254
7589 A Coupled Model for Two-Phase Simulation of a Heavy Water Pressure Vessel Reactor

Authors: D. Ramajo, S. Corzo, M. Nigro


A Multi-dimensional computational fluid dynamics (CFD) two-phase model was developed with the aim to simulate the in-core coolant circuit of a pressurized heavy water reactor (PHWR) of a commercial nuclear power plant (NPP). Due to the fact that this PHWR is a Reactor Pressure Vessel type (RPV), three-dimensional (3D) detailed modelling of the large reservoirs of the RPV (the upper and lower plenums and the downcomer) were coupled with an in-house finite volume one-dimensional (1D) code in order to model the 451 coolant channels housing the nuclear fuel. Regarding the 1D code, suitable empirical correlations for taking into account the in-channel distributed (friction losses) and concentrated (spacer grids, inlet and outlet throttles) pressure losses were used. A local power distribution at each one of the coolant channels was also taken into account. The heat transfer between the coolant and the surrounding moderator was accurately calculated using a two-dimensional theoretical model. The implementation of subcooled boiling and condensation models in the 1D code along with the use of functions for representing the thermal and dynamic properties of the coolant and moderator (heavy water) allow to have estimations of the in-core steam generation under nominal flow conditions for a generic fission power distribution. The in-core mass flow distribution results for steady state nominal conditions are in agreement with the expected from design, thus getting a first assessment of the coupled 1/3D model. Results for nominal condition were compared with those obtained with a previous 1/3D single-phase model getting more realistic temperature patterns, also allowing visualize low values of void fraction inside the upper plenum. It must be mentioned that the current results were obtained by imposing prescribed fission power functions from literature. Therefore, results are showed with the aim of point out the potentiality of the developed model.

Keywords: PHWR, CFD, thermo-hydraulic, two-phase flow

Procedia PDF Downloads 390
7588 Hydraulic Studies on Core Components of PFBR

Authors: G. K. Pandey, D. Ramadasu, I. Banerjee, V. Vinod, G. Padmakumar, V. Prakash, K. K. Rajan


Detailed thermal hydraulic investigations are very essential for safe and reliable functioning of liquid metal cooled fast breeder reactors. These investigations are further more important for components with complex profile, since there is no direct correlation available in literature to evaluate the hydraulic characteristics of such components directly. In those cases available correlations for similar profile or geometries may lead to significant uncertainty in the outcome. Hence experimental approach can be adopted to evaluate these hydraulic characteristics more precisely for better prediction in reactor core components. Prototype Fast Breeder Reactor (PFBR), a sodium cooled pool type reactor is under advanced stage of construction at Kalpakkam, India. Several components of this reactor core require hydraulic investigation before its usage in the reactor. These hydraulic investigations on full scale models, carried out by experimental approaches using water as simulant fluid are discussed in the paper.

Keywords: fast breeder reactor, cavitation, pressure drop, reactor components

Procedia PDF Downloads 381
7587 Corrosion Behavior of Fe-Ni-Cr and Zr Alloys in Supercritical Water Reactors

Authors: Igor Svishchev, Kashif Choudhry


Progress in advanced energy technologies is not feasible without understanding how engineering materials perform under extreme environmental conditions. The corrosion behaviour of Fe-Ni-Cr and Zr alloys has been systematically examined under high-temperature and supercritical water flow conditions. The changes in elemental release rate and dissolved gas concentration provide valuable insights into the mechanism of passivation by forming oxide films. A non-intrusive method for monitoring the extent of surface oxidation based on hydrogen release rate has been developed. This approach can be used for the on-line monitoring corrosion behavior of reactor materials without the need to interrupt the flow and remove corrosion coupons. Surface catalysed thermochemical reactions may generate sufficient hydrogen to have an effect on the accumulation of oxidizing species generated by radiolytic processes in the heat transport systems of the supercritical water cooled nuclear reactor.

Keywords: high-temperature corrosion, non-intrusive monitoring, reactor materials, supercritical water

Procedia PDF Downloads 59
7586 An Innovative Use of Flow Columns in Electrocoagulation Reactor to Control Water Temperature

Authors: Khalid S. Hashim, Andy Shaw, Rafid Alkhaddar, David Phipps, Ortoneda Pedrola


Temperature is an essential parameter in the electrocoagulation process (EC) as it governs the solubility of electrodes and the precipitates and the collision rate of particles in water being treated. Although it has been about 100 years since the EC technology was invented and applied in water and wastewater treatment, the effects of temperature on the its performance were insufficiently investigated. Thus, the present project aims to fill this gap by an innovative use of perforated flow columns in the designing of a new EC reactor (ECR1). The new reactor (ECR1) consisted of a Perspex made cylinder container supplied with a flow column consisted of perorated discoid electrodes that made from aluminium. The flow column has been installed vertically, half submerged in the water being treated, inside a plastic cylinder. The unsubmerged part of the flow column works as a radiator for the water being treated. In order to investigate the performance of ECR1; water samples with different initial temperatures (15, 20, 25, 30, and 35 °C) to the ECR1 for 20 min. Temperature of effluent water samples were measured using Hanna meter (Model: HI 98130). The obtained results demonstrated that the ECR1 reduced water temperature from 35, 30, and 25 °C to 24.6, 23.8, and 21.8 °C respectively. While low water temperature, 15 °C, increased slowly to reach 19.1 °C after 15 minutes and kept the same level till the end of the treatment period. At the same time, water sample with initial temperature of 20 °C showed almost a steady level of temperature along the treatment process, where the temperature increased negligibly from 20 to 20.1 °C after 20 minutes of treatment. In conclusion, ECR1 is able to control the temperature of water being treated around the room temperature even when the initial temperature was high (35 °C) or low (15 °C).

Keywords: electrocoagulation, flow column, treatment, water temperature

Procedia PDF Downloads 336
7585 Numerical Simulation of Ultraviolet Disinfection in a Water Reactor

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


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

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

Procedia PDF Downloads 179
7584 Reduction of Plutonium Production in Heavy Water Research Reactor: A Feasibility Study through Neutronic Analysis Using MCNPX2.6 and CINDER90 Codes

Authors: H. Shamoradifar, B. Teimuri, P. Parvaresh, S. Mohammadi


One of the main characteristics of Heavy Water Moderated Reactors is their high production of plutonium. This article demonstrates the possibility of reduction of plutonium and other actinides in Heavy Water Research Reactor. Among the many ways for reducing plutonium production in a heavy water reactor, in this research, changing the fuel from natural Uranium fuel to Thorium-Uranium mixed fuel was focused. The main fissile nucleus in Thorium-Uranium fuels is U-233 which would be produced after neutron absorption by Th-232, so the Thorium-Uranium fuels have some known advantages compared to the Uranium fuels. Due to this fact, four Thorium-Uranium fuels with different compositions ratios were chosen in our simulations; a) 10% UO2-90% THO2 (enriched= 20%); b) 15% UO2-85% THO2 (enriched= 10%); c) 30% UO2-70% THO2 (enriched= 5%); d) 35% UO2-65% THO2 (enriched= 3.7%). The natural Uranium Oxide (UO2) is considered as the reference fuel, in other words all of the calculated data are compared with the related data from Uranium fuel. Neutronic parameters were calculated and used as the comparison parameters. All calculations were performed by Monte Carol (MCNPX2.6) steady state reaction rate calculation linked to a deterministic depletion calculation (CINDER90). The obtained computational data showed that Thorium-Uranium fuels with four different fissile compositions ratios can satisfy the safety and operating requirements for Heavy Water Research Reactor. Furthermore, Thorium-Uranium fuels have a very good proliferation resistance and consume less fissile material than uranium fuels at the same reactor operation time. Using mixed Thorium-Uranium fuels reduced the long-lived α emitter, high radiotoxic wastes and the radio toxicity level of spent fuel.

Keywords: Heavy Water Reactor, Burn up, Minor Actinides, Neutronic Calculation

Procedia PDF Downloads 186
7583 Quantum Modelling of AgHMoO4, CsHMoO4 and AgCsMoO4 Chemistry in the Field of Nuclear Power Plant Safety

Authors: Mohamad Saab, Sidi Souvi


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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7582 Tribological Response of Self-Mated Zircaloy-4 under Varying Conditions

Authors: Bharat Kumar, Deepak Kumar, Vijay Chaudhry


Zirconium alloys are widely used for the core components of a pressurized heavy water reactor (PHWR) or Canada deuterium (CANDU) reactor due to their low neutron absorption cross-section and excellent mechanical properties. The components made of Zirconium alloys are subjected to flow-induced vibrations, resulting in fretting wear at the interface of; pressure tubes and bearing pads, pressure tubes and calandria tubes, and calandria tubes and Liquid injection shutdown system (LISS) nozzles. There is a need to explore the tribological response under such conditions. Present work simulates the contact between calandria tube and LISS nozzle of PHWR/CANDU reactor as cylinder-on-cylinder contact configuration. Reciprocating tribo-tests were conducted on Zircaloy-4 (Zr-4) under the self-mated condition at varying amplitude, frequency, and sliding time. To understand the active wear mechanism, worn surfaces were analyzed using Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy (EDS). The change in amplitude severely affects the wear than other factors. The wear mechanism transits from adhesion to abrasion with increasing test amplitude. The dominant wear mechanisms are micro-cutting and micro-plowing followed by delamination in some areas. However, the coefficient of friction has indifferent behaviors.

Keywords: zircaloy-4, tribology, calandria tube, LISS nozzle, PHWR

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7581 Intensification of Ethyl Esters Synthesis Using a Packed-Bed Tubular Reactor at Supercritical Conditions

Authors: Camila da Silva, Simone Belorte de Andrade, Vitor Augusto dos Santos Garcia, Vladimir Ferreira Cabral, J. Vladimir Oliveira Lúcio Cardozo-Filho


In the present study, the non-catalytic transesterification of soybean oil in continuous mode using supercritical ethanol were investigated. Experiments were performed in a packed-bed tubular reactor (PBTR) and variable studied were reaction temperature (523 K to 598 K), pressure (10 MPa to 20 MPa), oil to ethanol molar ratio (1:10 to 1:40) and water concentration (0 wt% to 10 wt% in ethanol). Results showed that ethyl esters yields obtained in the PBTR were higher (> 20 wt%) than those verified in a tubular reactor (TR), due to improved mass transfer conditions attained in the PBTR. Results demonstrated that temperature, pressure, oil to ethanol molar ratio and water concentration had a positive effect on fatty acid ethyl esters (FAEE) production in the experimental range investigated, with appreciable reaction yields (90 wt%) achieved at 598 K, 20 MPa, oil to ethanol molar ratio of 1:40 and 10 wt% of water concentration.

Keywords: packed bed reactor, ethyl esters, continuous process, catalyst-free process

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7580 Design Optimization of the Primary Containment Building of a Pressurized Water Reactor

Authors: M. Hossain, A. H. Khan, M. A. R. Sarkar


Primary containment structure is one of the five safety layers of a nuclear facility which is needed to be designed in such a manner that it can withstand the pressure and excessive radioactivity during accidental situations. It is also necessary to ensure minimization of cost with maximum possible safety in order to make the design economically feasible and attractive. This paper attempts to identify the optimum design conditions for primary containment structure considering both mechanical and radiation safety keeping the economic aspects in mind. This work takes advantage of commercial simulation software to identify the suitable conditions without the requirement of costly experiments. Generated data may be helpful for further studies.

Keywords: PWR, concrete containment, finite element approach, neutron attenuation, Von Mises stress

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7579 An Object-Oriented Modelica Model of the Water Level Swell during Depressurization of the Reactor Pressure Vessel of the Boiling Water Reactor

Authors: Rafal Bryk, Holger Schmidt, Thomas Mull, Ingo Ganzmann, Oliver Herbst


Prediction of the two-phase water mixture level during fast depressurization of the Reactor Pressure Vessel (RPV) resulting from an accident scenario is an important issue from the view point of the reactor safety. Since the level swell may influence the behavior of some passive safety systems, it has been recognized that an assumption which at the beginning may be considered as a conservative one, not necessary leads to a conservative result. This paper discusses outcomes obtained during simulations of the water dynamics and heat transfer during sudden depressurization of a vessel filled up to a certain level with liquid water under saturation conditions and with the rest of the vessel occupied by saturated steam. In case of the pressure decrease e.g. due to the main steam line break, the liquid water evaporates abruptly, being a reason thereby, of strong transients in the vessel. These transients and the sudden emergence of void in the region occupied at the beginning by liquid, cause elevation of the two-phase mixture. In this work, several models calculating the water collapse and swell levels are presented and validated against experimental data. Each of the models uses different approach to calculate void fraction. The object-oriented models were developed with the Modelica modelling language and the OpenModelica environment. The models represent the RPV of the Integral Test Facility Karlstein (INKA) – a dedicated test rig for simulation of KERENA – a new Boiling Water Reactor design of Framatome. The models are based on dynamic mass and energy equations. They are divided into several dynamic volumes in each of which, the fluid may be single-phase liquid, steam or a two-phase mixture. The heat transfer between the wall of the vessel and the fluid is taken into account. Additional heat flow rate may be applied to the first volume of the vessel in order to simulate the decay heat of the reactor core in a similar manner as it is simulated at INKA. The comparison of the simulations results against the reference data shows a good agreement.

Keywords: boiling water reactor, level swell, Modelica, RPV depressurization, thermal-hydraulics

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7578 A Study of Anoxic - Oxic Microbiological Technology for Treatment of Heavy Oily Refinery Wastewater

Authors: Di Wang, Li Fang, Shengyu Fang, Jianhua Li, Honghong Dong, Zhongzhi Zhang


Heavy oily refinery wastewater with the characteristics of high concentration of toxic organic pollutant, poor biodegradability and complicated dissolved recalcitrant compounds is intractable to be degraded. In order to reduce the concentrations of COD and total nitrogen pollutants which are the major pollutants in heavy oily refinery wastewater, the Anoxic - Oxic microbiological technology relies mainly on anaerobic microbial reactor which works with methanogenic archaea mainly that can convert organic pollutants to methane gas, and supplemented by aerobic treatment. The results of continuous operation for 2 months with a hydraulic retention time (HRT) of 60h showed that, the COD concentration from influent water of anaerobic reactor and effluent water from aerobic reactor were 547.8mg/L and 93.85mg/L, respectively. The total removal rate of COD was up to 84.9%. Compared with the 46.71mg/L of total nitrogen pollutants in influent water of anaerobic reactor, the concentration of effluent water of aerobic reactor decreased to 14.11mg/L. In addition, the average removal rate of total nitrogen pollutants reached as high as 69.8%. Based on the data displayed, Anoxic - Oxic microbial technology shows a great potential to dispose heavy oil sewage in energy saving and high-efficiency of biodegradation.

Keywords: anoxic - oxic microbiological technology, COD, heavy oily refinery wastewater, total nitrogen pollutant

Procedia PDF Downloads 375