The Experts below are selected from a list of 237 Experts worldwide ranked by ideXlab platform
Igor Pioro - One of the best experts on this subject based on the ideXlab platform.
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Heat Transfer to Supercritical Water (Liquid-Like State) Flowing in a Short Vertical Bare Tube With Upward Flow
Volume 9: Student Paper Competition, 2018Co-Authors: Alexander Zvorykin, Mohammed Mahdi, Roman Popov, K. Barati Far, Igor PioroAbstract:Current Nuclear Power Plants (NPPs) equipped with water-cooled Reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation Reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants. Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV Nuclear-Reactor Concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR Concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs. Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required. Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.
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Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat
Volume 5: Student Paper Competition, 2016Co-Authors: Yifeng Zhou, Paul Ponomaryov, Cristina Mazza, Igor PioroAbstract:Currently, i.e., in 2016, 4361 Nuclear-power Reactors operate in the world. 96.6% of these Reactors are water-cooled (373 Reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 Reactors — PHWRs are cooled with heavy water. 15% of all water-cooled Reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled Reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%). Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of Nuclear-power Reactors. It should be noted that among six Generation-IV Nuclear-Reactor Concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of Reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive Reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature. As an interim way of thermal-efficiency improvement for water-cooled NPPs Nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel Reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the Nuclear steam reheat in subcritical-pressure experimental boiling Reactors. Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel Reactors with Nuclear steam reheat. For this purpose the DE-TOP program has been used.
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Investigation of ATHLET System Code for Supercritical Water Applications
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: Jeffrey Samuel, Glenn Harvel, G. Lerchl, Igor PioroAbstract:SuperCritical Water-cooled Reactors (SCWRs) are one of six Generation-IV Nuclear-Reactor Concepts. They are expected to have high thermal efficiencies within the range of 45–50% owing to the Reactor’s high pressures and outlet temperatures. Efforts have been made to study the supercritical phenomena both analytically and experimentally. However, codes that have been used to study the phenomena analytically have not been validated for supercritical water.The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid- and transport-properties packages, thus enabling a transition from subcritical to supercritical fluid states. This extension needs to be validated using experimental data.In this work, the applicability of ATHLET code to predict supercritical-water behaviour in various heat-transfer conditions is assessed. Several well-known heat-transfer correlations for supercritical fluids are added to the code and applied for the first time in ATHLET simulations of experiments. A numerical model in ATHLET is created to represent an experimental test section and results for the heat transfer coefficient, bulk fluid temperature, and the tube inside-wall temperature are compared with the experimental data.The results from the ATHLET simulations are promising in the Normal and Enhanced Heat-Transfer Regimes. However, important phenomena such as Deteriorated Heat Transfer are currently not accurately predicted. While ATHLET can be used to develop preliminary design solutions for SCWRs, a significant effort in analysis of experimental work is required to make further advancements in the use of ATHLET for SCW applications.Copyright © 2014 by ASME
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Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Volume 6: Beyond Design Basis Events; Student Paper Competition, 2013Co-Authors: Wargha Peiman, Igor Pioro, Kamiel GabrielAbstract:SuperCritical Water-cooled Nuclear Reactor (SCWR) is one of the six Nuclear-Reactor Concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a Reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits.The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC Nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.Copyright © 2013 by ASME
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Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes
Volume 6: Beyond Design Basis Events; Student Paper Competition, 2013Co-Authors: Sahil Gupta, Eugene Saltanov, Igor PioroAbstract:Canada among many other countries is in pursuit of developing next generation (Generation IV) Nuclear-Reactor Concepts. One of the main objectives of Generation-IV Concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV Reactor design Concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions.To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations.Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data.In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.Copyright © 2013 by ASME
Ehud Greenspan - One of the best experts on this subject based on the ideXlab platform.
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phenomenology methods and experimental program for fluoride salt cooled high temperature Reactors fhrs
Progress in Nuclear Energy, 2014Co-Authors: Nicolas Zweibaum, Michael R Laufer, Charles Forsberg, G Cao, Mark H Anderson, Anselmo T Cisneros, Brian C Kelleher, Raluca O Scarlat, Jeffrey E Seifried, Ehud GreenspanAbstract:Abstract Due to their combination of high-temperature coated-particle fuel, molten salt coolant and related materials requirements, fluoride-salt-cooled, high-temperature Reactors (FHRs) exhibit different thermal hydraulic, neutronic and structural mechanics phenomena compared to conventional and more extensively studied other advanced Nuclear Reactor Concepts. This paper highlights key phenomena unique to FHRs, and reviews general issues for developing, verifying, and validating evaluation models for FHR technology that may apply to other advanced Reactors. System response codes that are appropriate to predict the behavior of FHRs under steady-state operation and licensing basis events are identified, along with experimental data needs to validate these codes. FHR materials requirements are highlighted, and the missions and licensing program for an FHR test Reactor, providing ultimate validation data and proof of concept before a commercial prototype is built, are presented. This review draws upon information compiled in a series of four white papers based on FHR experts workshops held in 2012 in the U.S.
Sundaresan Subramanian - One of the best experts on this subject based on the ideXlab platform.
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Design Considerations for Compact Ceramic Off-Set Strip Fin High Temperature Heat Exchangers
Volume 1: Turbo Expo 2005, 2005Co-Authors: Sundaresan Subramanian, Yitung Chen, Clayton Ray Delosier, Anthony Hechanova, Roald Akberov, Per F. PetersonAbstract:This paper deals with the development of a threedimensional numerical model to predict the overall performance of an advanced high temperature heat exchanger (HTHX) design, up to 1000 o C, for the production of hydrogen by the sulfur iodine thermo-chemical cycle used in advanced Nuclear Reactor Concepts. The design is an offset strip-fin, hybrid plate compact heat exchanger made from a liquid silicon impregnated carbon composite material. The two working fluids are helium gas and liquid salt (FLINAK). The offset strip-fin is chosen as a method of heat transfer enhancement because of its ability to induce periodic boundary layer restart mechanism between the fins that has a direct effect on heat transfer enhancement. The effects of the fin geometry on the flow field and heat transfer are studied in three-dimensions using Computational Fluid Dynamics (CFD) techniques, and the results are then compared with the results from the analytical calculations. The pre-processor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT that is based on the finite volume method is used to produce the numerical results. Fin dimensions need to be chosen that optimize heat transfer and minimize pressure drop. Comparisons of the overall performance between the rectangular and curved fin geometry were performed using computational fluid dynamics techniques. The model developed in this paper will be used to investigate the heat exchanger design parameters in order to find an optimal design. Also numerical simulation results were performed and compared to study the effect of the temperature dependent physical properties.
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The Effect of Fin Geometry on Design of Compact Off-Set Strip Fin High Temperature Heat Exchanger
Heat Transfer Part A, 2005Co-Authors: Sundaresan Subramanian, Valery Ponyavin, Yitung Chen, Clayton Ray De Losier, E. Hechanova, Per F. PetersonAbstract:This paper deals with the development of a three-dimensional numerical model to predict the overall performance of an advanced high temperature heat exchanger design, up to 1000°C, for the production of hydrogen by the sulfur iodine thermo-chemical cycle used in advanced Nuclear Reactor Concepts. The design is an offset strip-fin, hybrid plate compact heat exchanger made from a liquid silicon impregnated carbon composite material. The two working fluids are helium gas and molten salt (Flinak). The offset strip-fin is chosen as a method of heat transfer enhancement due to the boundary layer restart mechanism between the fins that has a direct effect on heat transfer enhancement. The effects of the fin geometry on the flow field and heat transfer are studied in three-dimensions using Computational Fluid Dynamics (CFD) techniques. The pre-processor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT that is based on the finite volume method is used to produce the numerical results. Fin dimensions need to be chosen that optimize heat transfer and minimize pressure drop. Comparison of the overall performance between two fin shapes (rectangular versus curved edges) is performed using computational fluid dynamics techniques. Fin and channel dimensions need to be chosen such as to optimize heat transfer performance and minimize pressure drop. The study is conducted with helium gas and liquid salt as the working fluids with a variety of Reynolds number values and fin dimensions. Both laminar and turbulent modeling is performed for the helium side fluid flow. The effect of the fin geometry is performed computational fluid dynamics techniques and optimization studies are performed. The model developed in this paper is used to investigate the heat exchanger design parameters in order to find an optimal design.Copyright © 2005 by ASME
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Development of an Advanced High Temperature Heat Exchanger Design for Hydrogen Production
Heat Transfer Volume 2, 2004Co-Authors: Sundaresan Subramanian, Yitung Chen, Anthony Hechanova, Roald Akberove, Clayton Ray De LosierAbstract:This paper deals with the development of an advanced high temperature heat exchanger design for hydrogen production by the sulfur iodine thermochemical cycle from advanced Nuclear Reactor Concepts. The offset strip-fin hybrid plate type compact heat exchanger concept is chosen, and the material of manufacture is the liquid silicon impregnated carbon composite. The offset strip-fin is chosen as a method of heat transfer enhancement due to the boundary layer restart mechanism between the fins that has a direct effect on enhancing heat transfer. The effect of the fin thickness, pitch in flow direction, and the aspect ratio of the offset fins on the flow field and heat transfer are studied in 2-D using Computational Fluid Dynamics (CFD) techniques, and the results are then compared with the analytical calculation results. The preprocessor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT, that is based on the finite volume method is used to produce numerical results. Proper dimensions of the strip fins need to be chosen in order to have an optimized heat transfer enhancement coupled with a reduced pressure drop. The study is conducted with helium gas as the working fluid with varied of Reynolds number values. The flow and heat transfer is considered to become periodically fully developed after a certain entrance length hence numerical simulations were performed using periodic boundary conditions. Two-dimensional numerical simulations were also performed for the whole length of the heat exchanger which has 37 such periodic modules. Comparison study was performed between the cases of fins with rectangular and curved geometry. Attempt has also been made in order to validate the coefficient of fin thickness (Cfin) value using CFD techniques, which has been used in the existing empirical correlations to suit this type of heat exchanger geometry. The model developed in this paper is used to investigate the heat exchanger design parameters in order to find an optimal design.
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Development of an Advanced High Temperature Heat Exchanger Design for Hydrogen Production
Heat Transfer Volume 2, 2004Co-Authors: Sundaresan Subramanian, Yitung Chen, Anthony Hechanova, Roald Akberove, Clayton Ray De LosierAbstract:This paper deals with the development of an advanced high temperature heat exchanger design for hydrogen production by the sulfur iodine thermochemical cycle from advanced Nuclear Reactor Concepts. The offset strip-fin hybrid plate type compact heat exchanger concept is chosen, and the material of manufacture is the liquid silicon impregnated carbon composite. The offset strip-fin is chosen as a method of heat transfer enhancement due to the boundary layer restart mechanism between the fins that has a direct effect on enhancing heat transfer. The effect of the fin thickness, pitch in flow direction, and the aspect ratio of the offset fins on the flow field and heat transfer are studied in 2-D using Computational Fluid Dynamics (CFD) techniques, and the results are then compared with the analytical calculation results. The preprocessor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT, that is based on the finite volume method is used to produce numerical results. Proper dimensions of the strip fins need to be chosen in order to have an optimized heat transfer enhancement coupled with a reduced pressure drop. The study is conducted with helium gas as the working fluid with varied of Reynolds number values. The flow and heat transfer is considered to become periodically fully developed after a certain entrance length hence numerical simulations were performed using periodic boundary conditions. Two-dimensional numerical simulations were also performed for the whole length of the heat exchanger which has 37 such periodic modules. Comparison study was performed between the cases of fins with rectangular and curved geometry. Attempt has also been made in order to validate the coefficient of fin thickness (Cfin ) value using CFD techniques, which has been used in the existing empirical correlations to suit this type of heat exchanger geometry. The model developed in this paper is used to investigate the heat exchanger design parameters in order to find an optimal design.Copyright © 2004 by ASME
Wargha Peiman - One of the best experts on this subject based on the ideXlab platform.
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Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors
Journal of Nuclear Engineering and Radiation Science, 2015Co-Authors: Wargha Peiman, K. GabrielAbstract:To address the need to develop new Nuclear Reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of Nuclear Reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the Nuclear Reactors to six Concepts, one of which is supercritical water-cooled Reactor (SCWR). Among the Generation IV Nuclear-Reactor Concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled Reactors (pressurized water Reactors (PWRs), boiling water Reactors (BWRs), pressurized heavy water Reactors (PHWRs), and light-water, graphite-moderated Reactors (LGRs)), which comprise 96% of the current fleet of operating Nuclear power Reactors and are categorized under Generation II, III, and III+ Nuclear Reactors. The latter water-cooled Reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the Reactor core should be designed, which depends on a fuel-channel design. Second, a Nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel Concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.
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Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Volume 6: Beyond Design Basis Events; Student Paper Competition, 2013Co-Authors: Wargha Peiman, Igor Pioro, Kamiel GabrielAbstract:SuperCritical Water-cooled Nuclear Reactor (SCWR) is one of the six Nuclear-Reactor Concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a Reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits.The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC Nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.Copyright © 2013 by ASME
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Sensitivity Analysis of Fuel Centerline Temperature in SCWRs
Volume 5: Fusion Engineering; Student Paper Competition; Design Basis and Beyond Design Basis Events; Simple and Combined Cycles, 2012Co-Authors: Ayman Abdalla, Wargha Peiman, Igor Pioro, Kamiel GabrielAbstract:The Generation IV International Forum (GIF) is intended to encourage the world’s leading Nuclear countries to develop Nuclear energy systems that can supply future energy demands. There are six Nuclear Reactor Concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six Nuclear-Reactor Concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional Nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%.The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO2+BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MWth.Copyright © 2012 by ASME
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Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)
Volume 5: Fusion Engineering; Student Paper Competition; Design Basis and Beyond Design Basis Events; Simple and Combined Cycles, 2012Co-Authors: Wargha Peiman, Igor Pioro, Eu. Saltanov, Lisa Grande, B. Rouben, Kamiel GabrielAbstract:SuperCritical Water-cooled Nuclear Reactor (SCWR) designs are one of six Nuclear-Reactor Concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such Reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the Cu–Cl cycle.The main objective of this paper is to determine the power distribution inside the core of an SCWR by using a lattice code - DRAGON and a diffusion code - DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermal-hydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature for UO2 and UC Nuclear fuels. Results of an analysis showed that the fuel centerline temperature of UC was significantly lower than that of UO2. This paper also studies effects of energy groups on multi-group diffusion calculations and proposes nine energy groups for further neutronic studies related to SCWRs.Copyright © 2012 by ASME
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Steam-Reheat Options for Pressure-Tube SCWRs
18th International Conference on Nuclear Engineering: Volume 2, 2010Co-Authors: Eugene Saltanov, Wargha Peiman, Maria Naidin, Amjad Farah, Krysten King, Igor PioroAbstract:Concepts of Nuclear Reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 40-year break, the idea of developing Nuclear Reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in Nuclear Reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30–35% to about 45–48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency Nuclear Steam Reheat (NSR) has to be introduced inside a Reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the Nuclear steam reheat at subcritical-pressure experimental boiling Reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW Nuclear Reactor Concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the Nuclear steam reheat at subcritical pressures in pressure-tube Reactors is easier task than that in pressure-vessel Reactors. Some design features related to the NSR are discussed in this paper. The main conclusion is that the development of SCW pressure-tube Nuclear Reactors with the Nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.Copyright © 2010 by ASME
Per F. Peterson - One of the best experts on this subject based on the ideXlab platform.
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The Effect of Fin Geometry on Design of Compact Off-Set Strip Fin High Temperature Heat Exchanger
Heat Transfer Part A, 2005Co-Authors: Sundaresan Subramanian, Valery Ponyavin, Yitung Chen, Clayton Ray De Losier, E. Hechanova, Per F. PetersonAbstract:This paper deals with the development of a three-dimensional numerical model to predict the overall performance of an advanced high temperature heat exchanger design, up to 1000°C, for the production of hydrogen by the sulfur iodine thermo-chemical cycle used in advanced Nuclear Reactor Concepts. The design is an offset strip-fin, hybrid plate compact heat exchanger made from a liquid silicon impregnated carbon composite material. The two working fluids are helium gas and molten salt (Flinak). The offset strip-fin is chosen as a method of heat transfer enhancement due to the boundary layer restart mechanism between the fins that has a direct effect on heat transfer enhancement. The effects of the fin geometry on the flow field and heat transfer are studied in three-dimensions using Computational Fluid Dynamics (CFD) techniques. The pre-processor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT that is based on the finite volume method is used to produce the numerical results. Fin dimensions need to be chosen that optimize heat transfer and minimize pressure drop. Comparison of the overall performance between two fin shapes (rectangular versus curved edges) is performed using computational fluid dynamics techniques. Fin and channel dimensions need to be chosen such as to optimize heat transfer performance and minimize pressure drop. The study is conducted with helium gas and liquid salt as the working fluids with a variety of Reynolds number values and fin dimensions. Both laminar and turbulent modeling is performed for the helium side fluid flow. The effect of the fin geometry is performed computational fluid dynamics techniques and optimization studies are performed. The model developed in this paper is used to investigate the heat exchanger design parameters in order to find an optimal design.Copyright © 2005 by ASME
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Design Considerations for Compact Ceramic Off-Set Strip Fin High Temperature Heat Exchangers
Volume 1: Turbo Expo 2005, 2005Co-Authors: Sundaresan Subramanian, Yitung Chen, Clayton Ray Delosier, Anthony Hechanova, Roald Akberov, Per F. PetersonAbstract:This paper deals with the development of a threedimensional numerical model to predict the overall performance of an advanced high temperature heat exchanger (HTHX) design, up to 1000 o C, for the production of hydrogen by the sulfur iodine thermo-chemical cycle used in advanced Nuclear Reactor Concepts. The design is an offset strip-fin, hybrid plate compact heat exchanger made from a liquid silicon impregnated carbon composite material. The two working fluids are helium gas and liquid salt (FLINAK). The offset strip-fin is chosen as a method of heat transfer enhancement because of its ability to induce periodic boundary layer restart mechanism between the fins that has a direct effect on heat transfer enhancement. The effects of the fin geometry on the flow field and heat transfer are studied in three-dimensions using Computational Fluid Dynamics (CFD) techniques, and the results are then compared with the results from the analytical calculations. The pre-processor GAMBIT is used to create a computational mesh, and the CFD software package FLUENT that is based on the finite volume method is used to produce the numerical results. Fin dimensions need to be chosen that optimize heat transfer and minimize pressure drop. Comparisons of the overall performance between the rectangular and curved fin geometry were performed using computational fluid dynamics techniques. The model developed in this paper will be used to investigate the heat exchanger design parameters in order to find an optimal design. Also numerical simulation results were performed and compared to study the effect of the temperature dependent physical properties.
