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Joo Hwan Park - One of the best experts on this subject based on the ideXlab platform.
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Development and validation of a CATHENA Fuel Channel model for a post-blowdown analysis of the high temperature thermal–chemical experiment CS28-1
Annals of Nuclear Energy, 2009Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for a new methodology for a Fuel Channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown Fuel Channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H 2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the Fuel element simulators (FESs) for the three Fuel rings and the pressure tube were reasonably good as proven by the simulation results. This is thought to be due to success in reproducing the initial steady state condition well, and due to the detailed modeling features of CATHENA code for the coupled conduction, convection, and radiation heat transfer in a complex Fuel bundle geometry. However, one problem still remained unresolved, i.e. the inability to accurately predict the pressure tube temperature at the initial steady state condition, and an adjustment of the CO 2 gap conductivity necessary to match the measured pressure tube temperatures. However, this raises a question as to how the transient FES and the pressure tube temperature can be predicted so well in spite of an insufficient justification for using the “non-participating medium assumption” for the CO 2 gas gap. Through this study and the affiliated previous study for the steady state, it was found that the radiation heat transfer model of CATHENA among the FESs of the three rings and the pressure tube as well as the exothermic metal–water reaction model based on the Urbanic–Heidrick correlation are reasonably accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for an accurate prediction of the whole transient as it serves as the starting point for the transient.
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Validation of a CATHENA Fuel Channel model for the post blowdown analysis of the high temperature thermal–chemical experiment CS28-1, I – Steady state
Annals of Nuclear Energy, 2008Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for the new methodology of the Fuel Channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown Fuel Channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown Fuel Channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a Fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the Fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20–25 °C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using “transparent” assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code’s inherent modeling feature is raised from this study. Based on a comparison of the prediction of the current CATHENA model and the experiment data, the steady state model is deemed to be adequate as a starting point for the following high temperature thermal–chemical experiment of a metal–water reaction.
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Application of a Zircaloy/Steam Oxidation Model to a CFD Code and its Validation against a CANDU Fuel Channel Experiment: CS28-2
Journal of Nuclear Science and Technology, 2007Co-Authors: Hyoung Tae Kim, Bo Wook Rhee, Joo Hwan ParkAbstract:Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU Fuel Channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU Fuel Channel.
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Benchmark Calculations of a Radiation Heat Transfer for a CANDU Fuel Channel Analysis using the CFD Code
Journal of Nuclear Science and Technology, 2006Co-Authors: Hyoung Tae Kim, Bo Wook Rhee, Joo Hwan ParkAbstract:To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU Fuel Channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU Fuel Channel configurations from a simple geometry to a whole Fuel Channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is ...
Bo Wook Rhee - One of the best experts on this subject based on the ideXlab platform.
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Coupled neutronics/thermal–hydraulics analysis of CANDU–SCWR Fuel Channel
Annals of Nuclear Energy, 2010Co-Authors: Jianqiang Shan, Bo Wook Rhee, Wei Chen, Laurence K.h. LeungAbstract:Abstract At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subChannel code Advanced Thermal–Hydraulics Analysis SubChannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR Fuel Channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR Fuel bundles; improved radial Fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.
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coupled neutronics thermal hydraulics analysis of candu scwr Fuel Channel
Annals of Nuclear Energy, 2010Co-Authors: Jianqiang Shan, Bo Wook Rhee, Wei Chen, Laurence K.h. LeungAbstract:Abstract At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subChannel code Advanced Thermal–Hydraulics Analysis SubChannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR Fuel Channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR Fuel bundles; improved radial Fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.
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Development and validation of a CATHENA Fuel Channel model for a post-blowdown analysis of the high temperature thermal–chemical experiment CS28-1
Annals of Nuclear Energy, 2009Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for a new methodology for a Fuel Channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown Fuel Channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H 2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the Fuel element simulators (FESs) for the three Fuel rings and the pressure tube were reasonably good as proven by the simulation results. This is thought to be due to success in reproducing the initial steady state condition well, and due to the detailed modeling features of CATHENA code for the coupled conduction, convection, and radiation heat transfer in a complex Fuel bundle geometry. However, one problem still remained unresolved, i.e. the inability to accurately predict the pressure tube temperature at the initial steady state condition, and an adjustment of the CO 2 gap conductivity necessary to match the measured pressure tube temperatures. However, this raises a question as to how the transient FES and the pressure tube temperature can be predicted so well in spite of an insufficient justification for using the “non-participating medium assumption” for the CO 2 gas gap. Through this study and the affiliated previous study for the steady state, it was found that the radiation heat transfer model of CATHENA among the FESs of the three rings and the pressure tube as well as the exothermic metal–water reaction model based on the Urbanic–Heidrick correlation are reasonably accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for an accurate prediction of the whole transient as it serves as the starting point for the transient.
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Validation of a CATHENA Fuel Channel model for the post blowdown analysis of the high temperature thermal–chemical experiment CS28-1, I – Steady state
Annals of Nuclear Energy, 2008Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for the new methodology of the Fuel Channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown Fuel Channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown Fuel Channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a Fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the Fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20–25 °C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using “transparent” assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code’s inherent modeling feature is raised from this study. Based on a comparison of the prediction of the current CATHENA model and the experiment data, the steady state model is deemed to be adequate as a starting point for the following high temperature thermal–chemical experiment of a metal–water reaction.
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Application of a Zircaloy/Steam Oxidation Model to a CFD Code and its Validation against a CANDU Fuel Channel Experiment: CS28-2
Journal of Nuclear Science and Technology, 2007Co-Authors: Hyoung Tae Kim, Bo Wook Rhee, Joo Hwan ParkAbstract:Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU Fuel Channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU Fuel Channel.
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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Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: Wargha Peiman, Igor Pioro, Kamiel 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 the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%.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 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the 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 (for PCh SCWR). Second, a nuclear Fuel and Fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results.Several Fuel-Channel designs have been proposed for SCWRs. These Fuel-Channel designs can be classified into two categories: direct-flow and re-entrant Channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant Fuel Channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the Fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles.A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a Channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the Fuel centerline temperature of a newly designed Fuel bundle with UO2 as a reference Fuel. The results show that the maximum Fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for Fuel and sheath, respectively.Copyright © 2014 by ASME
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Pressure Drop Analysis of a Re-Entrant Fuel Channel in a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: S. Maghsoudi, Wargha Peiman, Igor Pioro, K. GabrielAbstract:Pressure drop calculation and temperature profiles associated with Fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a Fuel Channel of a SuperCritical Water-cooled Reactor (SCWR). One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions. The total pressure drop due to all these parameters was between 108 and 121 kPa.
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Pressure Drop Analysis of a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Volume 15: Safety Reliability and Risk; Virtual Podium (Posters), 2013Co-Authors: A. Dragunov, Wargha PeimanAbstract:Pressure drop calculation and temperature profiles associated with Fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a Fuel Channel of a SuperCritical Water-cooled Reactor (SCWR) and to calculate the temperature profile of the sheath and the Fuel bundles. One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions.Copyright © 2013 by ASME
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Heat-Loss Calculations in a SCWR Fuel-Channel
18th International Conference on Nuclear Engineering: Volume 2, 2010Co-Authors: Wargha Peiman, Eugene Saltanov, Kamiel Gabriel, Igor PioroAbstract:The objective of this paper is to calculate heat losses from a CANDU-6 Fuel-Channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from Fuel-Channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the Fuel-Channel were calculated along the heated length of the Fuel-Channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The Fuel-Channel was divided into small segments and for each segment thermal resistances of the Fuel-Channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per Fuel-Channel when the pressure of the annulus gas was 0.3 MPa.Copyright © 2010 by ASME
Igor Pioro - One of the best experts on this subject based on the ideXlab platform.
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Pressure Drop Analysis of a Re-Entrant Fuel Channel in a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: S. Maghsoudi, Wargha Peiman, Igor Pioro, K. GabrielAbstract:Pressure drop calculation and temperature profiles associated with Fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a Fuel Channel of a SuperCritical Water-cooled Reactor (SCWR). One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions. The total pressure drop due to all these parameters was between 108 and 121 kPa.
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Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: Wargha Peiman, Igor Pioro, Kamiel 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 the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%.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 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the 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 (for PCh SCWR). Second, a nuclear Fuel and Fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results.Several Fuel-Channel designs have been proposed for SCWRs. These Fuel-Channel designs can be classified into two categories: direct-flow and re-entrant Channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant Fuel Channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the Fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles.A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a Channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the Fuel centerline temperature of a newly designed Fuel bundle with UO2 as a reference Fuel. The results show that the maximum Fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for Fuel and sheath, respectively.Copyright © 2014 by ASME
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Materials and Stress Analysis of Fuel-Channel of a Generic 1200-MWel Pressure-Channel Reactor With Nuclear Steam Reheat
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: Rand Abdullah, Matthew Baldock, Andrei Vincze, Khalil Sidawi, Igor PioroAbstract:The objective of this paper is to modify the current existing CANFLEX® Fuel-bundle design to examine its ability to withstand high-temperature conditions of a proposed generic reactor with nuclear steam reheat. One of this reactor’s characteristics is having Super-Heated Steam (SHS) Channels in addition to Pressurized-Water (PW) Channels in order to increase the thermal efficiency of the plant by about 7–12%. This increase may be attained by raising the outlet temperature of the SHS-Channels coolant to about 550°C. Operating at the higher temperatures will definitely have an effect on the mechanical and neutronic properties of Fuel-Channel materials, specifically on Fuel-sheath and pressure-tube materials.This paper compares Inconel-600 and SS-304 in order to determine the most suitable material for SHS-Channel’s sheath and pressure tube. This is achieved by comparing strength of materials by performing stress- and displacement-analysis simulation using NX8.5 software (NX8.5, 2009).The analysis in this paper can also be applied to other Nuclear-Power Plants (NPPs) that require operating at higher temperatures such as Super-Critical Water-cooled Reactors (SCWRs).Copyright © 2014 by ASME
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Heat-Loss Calculations in a SCWR Fuel-Channel
18th International Conference on Nuclear Engineering: Volume 2, 2010Co-Authors: Wargha Peiman, Eugene Saltanov, Kamiel Gabriel, Igor PioroAbstract:The objective of this paper is to calculate heat losses from a CANDU-6 Fuel-Channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from Fuel-Channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the Fuel-Channel were calculated along the heated length of the Fuel-Channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The Fuel-Channel was divided into small segments and for each segment thermal resistances of the Fuel-Channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per Fuel-Channel when the pressure of the annulus gas was 0.3 MPa.Copyright © 2010 by ASME
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Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel
Volume 4: Codes Standards Licensing and Regulatory Issues; Student Paper Competition, 2009Co-Authors: Leyland Allison, Lisa Grande, Sally Mikhael, Adrianexy Rodriguez Prado, Bryan Villamere, Igor PioroAbstract:SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear Fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2 ) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower Fuel centerline temperature compared to those of conventional nuclear Fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the Fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR Fuel Channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and Fuel centreline and HTC profiles were calculated along the heated length of a Fuel Channel.Copyright © 2009 by ASME
Hyoung Tae Kim - One of the best experts on this subject based on the ideXlab platform.
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Development and validation of a CATHENA Fuel Channel model for a post-blowdown analysis of the high temperature thermal–chemical experiment CS28-1
Annals of Nuclear Energy, 2009Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for a new methodology for a Fuel Channel safety analysis code for CANDU-6 nuclear reactor, a CATHENA model for a post-blowdown Fuel Channel analysis has been developed, and tested for a high temperature thermal–chemical experiment CS28-1 [Lei, Q.M., 1993. Post-test analysis of the 28-element high-temperature thermal–chemical experiment CS28-1. In: 4th International Conference on Simulation Methods in Nuclear Engineering, Montreal, PQ, 1993]. Pursuant to the objective of this investigation, the current study has focused on understanding the involved phenomena, their interrelations, and how to maintain a good accuracy of the temperature and H 2 generation rate prediction without losing the important physics of the involved phenomena. The transient simulation results for the Fuel element simulators (FESs) for the three Fuel rings and the pressure tube were reasonably good as proven by the simulation results. This is thought to be due to success in reproducing the initial steady state condition well, and due to the detailed modeling features of CATHENA code for the coupled conduction, convection, and radiation heat transfer in a complex Fuel bundle geometry. However, one problem still remained unresolved, i.e. the inability to accurately predict the pressure tube temperature at the initial steady state condition, and an adjustment of the CO 2 gap conductivity necessary to match the measured pressure tube temperatures. However, this raises a question as to how the transient FES and the pressure tube temperature can be predicted so well in spite of an insufficient justification for using the “non-participating medium assumption” for the CO 2 gas gap. Through this study and the affiliated previous study for the steady state, it was found that the radiation heat transfer model of CATHENA among the FESs of the three rings and the pressure tube as well as the exothermic metal–water reaction model based on the Urbanic–Heidrick correlation are reasonably accurate and sound. Also it was found that an accurate prediction of the initial condition of the experiment is very important for an accurate prediction of the whole transient as it serves as the starting point for the transient.
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Validation of a CATHENA Fuel Channel model for the post blowdown analysis of the high temperature thermal–chemical experiment CS28-1, I – Steady state
Annals of Nuclear Energy, 2008Co-Authors: Bo Wook Rhee, Hyoung Tae Kim, Joo Hwan ParkAbstract:Abstract To form a licensing basis for the new methodology of the Fuel Channel safety analysis code system for CANDU-6, a CATHENA model for the post-blowdown Fuel Channel analysis for a Large Break LOCA has been developed, and tested for the steady state of a high temperature thermal–chemical experiment CS28-1. As the major concerns of the post-blowdown Fuel Channel analysis of the current CANDU-6 design are how much of the decay heat can be discharged to the moderator via a radiation and a convective heat transfer at the expected accident conditions, and how much zirconium sheath would be oxidized to generate H2 at how high a Fuel temperature, this study has focused on understanding these phenomena, their interrelations, and a way to maintain a good accuracy in the prediction of the Fuel and the pressure tube temperatures without losing the important physics of the involved phenomena throughout the post-blowdown phase of a LBLOCA. For a better prediction, those factors that may significantly contribute to the prediction accuracy of the steady state of the test bundles were sought. The result shows that once the pressure tube temperature is predicted correctly by the CATHENA heat transfer model between the pressure tube and the calandria tube through a gap thermal resistance adjustment, all the remaining temperatures of the inner ring, middle ring and outer ring FES temperatures can be predicted quite satisfactorily, say to within an accuracy range of 20–25 °C, which is comparable to the reported accuracy of the temperature measurement, ±2%. Also the analysis shows the choice of the emissivity of the solid structures (typically, 0.80, 0.34, 0.34 for FES, PT, CT), and the thermal resistance across the CO2 annulus are factors that significantly affect the steady state prediction accuracy. A question on the legitimacy of using “transparent” assumption for the CO2 gas annulus for the radiation heat transfer between the pressure tube and the calandria tube in CATHENA code’s inherent modeling feature is raised from this study. Based on a comparison of the prediction of the current CATHENA model and the experiment data, the steady state model is deemed to be adequate as a starting point for the following high temperature thermal–chemical experiment of a metal–water reaction.
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Application of a Zircaloy/Steam Oxidation Model to a CFD Code and its Validation against a CANDU Fuel Channel Experiment: CS28-2
Journal of Nuclear Science and Technology, 2007Co-Authors: Hyoung Tae Kim, Bo Wook Rhee, Joo Hwan ParkAbstract:Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU Fuel Channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU Fuel Channel.
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Benchmark Calculations of a Radiation Heat Transfer for a CANDU Fuel Channel Analysis using the CFD Code
Journal of Nuclear Science and Technology, 2006Co-Authors: Hyoung Tae Kim, Bo Wook Rhee, Joo Hwan ParkAbstract:To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU Fuel Channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU Fuel Channel configurations from a simple geometry to a whole Fuel Channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with these solutions, so it is ...
