The Experts below are selected from a list of 288 Experts worldwide ranked by ideXlab platform
Wenxi Tian - One of the best experts on this subject based on the ideXlab platform.
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Development of a subchannel Analysis Code and its application to annular fuel assemblies
Annals of Nuclear Energy, 2019Co-Authors: Hang Xia, Wenxi Tian, Suizheng QiuAbstract:Abstract A Subchannel Analysis Code for Annular Fuel (SACAF) capable of modeling dual-cooled annular fuel pins was developed. The coolant flow distribution was adjusted until a uniform pressure drops was formed in all the subchannels. The heat transfer fraction was determined by the criterion that the temperature of the fuel pins was continuous and a burnup-dependent fuel thermal conductivity model was incorporated into the Code to analyze the effect of the thermal conductivity degradation. In this paper, the characteristics and verification of the SACAF are described. The thermal-hydraulic performance of the annular fuel was determined for the major thermal-hydraulic parameters and was compared with that of the cylindrical fuel assembly. For the same power distribution, the annular fuel exhibited a greater departure from nucleate boiling ratio (DNBR) and a much lower fuel pellet temperature than the cylindrical fuel. The fuel pin temperature profiles at different burnup levels were obtained and the effects of the thermal conductivity at different burnup levels on the temperature profile and the heat split were determined.
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Development and Verification of Thermal Hydraulic Subchannel Analysis Code for Motion Conditions
Volume 4: Computational Fluid Dynamics (CFD) and Coupled Codes; Decontamination and Decommissioning Radiation Protection Shielding and Waste Managemen, 2016Co-Authors: Rong Cai, Wenxi Tian, Siyang Huang, Kai Wang, Suizheng QiuAbstract:As the conventional core Analysis Codes are designed for the land-based reactor core, a thermal-hydraulic subchannel Analysis Code for motion conditions (SACROM) is developed. To evaluate the effect of different motion conditions on coolant flow, the model of additional forces is established. To check the accuracy of the models, the Code has been verified by test data, a commercial subchannel Code and a CFD Code. In the steady-state verification, the ISPRA data were used and the predicted results agree well with the test data. For the transient simulations without motion conditions, the Code COBRA-EN was chosen and the results from SACROM fit the results from COBRA-EN well. And CFX Code was used to verify the accuracy of the model of additional forces for motion conditions. The results show that the Code can be used in the thermal hydraulic characteristics of the reactor core under motion conditions.
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development and preliminary validation of a steam generator 3d thermohydraulics Analysis Code staf
Nuclear Engineering and Design, 2016Co-Authors: Wenxi Tian, Tenglong Cong, Rui Zhang, G H SuAbstract:Abstract Porous media model in Fluent Code, coupled with two-phase mixture flow model, resistance model of tubes and heat transfer model through tubes, is employed to develop a steam generator thermohydraulics Analysis Code STAF (Steam generator Thermohydraulics Analysis Code based on Fluent). In this Code, the heat transfer from primary to secondary side is calculated three-dimensionally during iteration. The localized velocity, temperature, enthalpy, quality and void fraction in steam generator can be obtained by this Code. STAF is validated in two ways. First, STAF is used to calculate the thermal-hydraulic parameters in steam generator of AP 1000. The calculated results are compared with designed values to prove that the coupled heat transfer calculation in STAF is accurate. Second, STAF is employed to simulate the FRIGG test to validate the localized parameter calculation performance by comparing the calculated localized void fraction with test values.
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the development of module in vessel degraded severe accident Analysis Code midac and the relevant research for cpr1000 during the station blackout scenario
Progress in Nuclear Energy, 2014Co-Authors: Jun Wang, Wenxi Tian, Yapei Zhang, Lie Chen, Luteng Zhang, Yukun Zhou, Suizheng QiuAbstract:Abstract To meet the domestic demand of software autonomous scheduling, Xi'an Jiaotong University had developed MIDAC (a Module In-vessel degraded severe accident Analysis Code) which can not only analyze the processes of in-vessel severe accident, but also provide the related results of each one. This Code was composed of five modules: the early behavior module, the core degradation module, the debris bed module, the melting materials IVR module and the connecting module. In this paper, the basic mathematic-physical models of those modules were briefly introduced. Then, the CPR1000 station blackout scenario severe accident was set as an example to calculate the primary system thermal-hydraulic transient, the core degradation and the debris behavior, as well as the creep rupture of heat structures in primary loop pressure boundary. At last, the accuracy of MIDAC was verified in partial comparison with SCDAP/RELAP5 in primary system thermal-hydraulic transient analyzing part.
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Development and Application of a UTSG Thermal-Hydraulic Analysis Code
Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle Radioactive Waste Management and Decommissioning; Computational Fluid D, 2014Co-Authors: Tenglong Cong, Wenxi Tian, Suizheng QiuAbstract:Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the Analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics Analysis Code, named STAF, is developed based on FLUENT. With STAF Code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This Code is developed based on the porous media theory. In this Code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF Code. The calculated results are compared with a general steam generator thermohydraulic Analysis Code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF Code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.
Suizheng Qiu - One of the best experts on this subject based on the ideXlab platform.
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Development of a subchannel Analysis Code and its application to annular fuel assemblies
Annals of Nuclear Energy, 2019Co-Authors: Hang Xia, Wenxi Tian, Suizheng QiuAbstract:Abstract A Subchannel Analysis Code for Annular Fuel (SACAF) capable of modeling dual-cooled annular fuel pins was developed. The coolant flow distribution was adjusted until a uniform pressure drops was formed in all the subchannels. The heat transfer fraction was determined by the criterion that the temperature of the fuel pins was continuous and a burnup-dependent fuel thermal conductivity model was incorporated into the Code to analyze the effect of the thermal conductivity degradation. In this paper, the characteristics and verification of the SACAF are described. The thermal-hydraulic performance of the annular fuel was determined for the major thermal-hydraulic parameters and was compared with that of the cylindrical fuel assembly. For the same power distribution, the annular fuel exhibited a greater departure from nucleate boiling ratio (DNBR) and a much lower fuel pellet temperature than the cylindrical fuel. The fuel pin temperature profiles at different burnup levels were obtained and the effects of the thermal conductivity at different burnup levels on the temperature profile and the heat split were determined.
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Development and Verification of Thermal Hydraulic Subchannel Analysis Code for Motion Conditions
Volume 4: Computational Fluid Dynamics (CFD) and Coupled Codes; Decontamination and Decommissioning Radiation Protection Shielding and Waste Managemen, 2016Co-Authors: Rong Cai, Wenxi Tian, Siyang Huang, Kai Wang, Suizheng QiuAbstract:As the conventional core Analysis Codes are designed for the land-based reactor core, a thermal-hydraulic subchannel Analysis Code for motion conditions (SACROM) is developed. To evaluate the effect of different motion conditions on coolant flow, the model of additional forces is established. To check the accuracy of the models, the Code has been verified by test data, a commercial subchannel Code and a CFD Code. In the steady-state verification, the ISPRA data were used and the predicted results agree well with the test data. For the transient simulations without motion conditions, the Code COBRA-EN was chosen and the results from SACROM fit the results from COBRA-EN well. And CFX Code was used to verify the accuracy of the model of additional forces for motion conditions. The results show that the Code can be used in the thermal hydraulic characteristics of the reactor core under motion conditions.
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the development of module in vessel degraded severe accident Analysis Code midac and the relevant research for cpr1000 during the station blackout scenario
Progress in Nuclear Energy, 2014Co-Authors: Jun Wang, Wenxi Tian, Yapei Zhang, Lie Chen, Luteng Zhang, Yukun Zhou, Suizheng QiuAbstract:Abstract To meet the domestic demand of software autonomous scheduling, Xi'an Jiaotong University had developed MIDAC (a Module In-vessel degraded severe accident Analysis Code) which can not only analyze the processes of in-vessel severe accident, but also provide the related results of each one. This Code was composed of five modules: the early behavior module, the core degradation module, the debris bed module, the melting materials IVR module and the connecting module. In this paper, the basic mathematic-physical models of those modules were briefly introduced. Then, the CPR1000 station blackout scenario severe accident was set as an example to calculate the primary system thermal-hydraulic transient, the core degradation and the debris behavior, as well as the creep rupture of heat structures in primary loop pressure boundary. At last, the accuracy of MIDAC was verified in partial comparison with SCDAP/RELAP5 in primary system thermal-hydraulic transient analyzing part.
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Development and Application of a UTSG Thermal-Hydraulic Analysis Code
Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle Radioactive Waste Management and Decommissioning; Computational Fluid D, 2014Co-Authors: Tenglong Cong, Wenxi Tian, Suizheng QiuAbstract:Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the Analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics Analysis Code, named STAF, is developed based on FLUENT. With STAF Code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This Code is developed based on the porous media theory. In this Code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF Code. The calculated results are compared with a general steam generator thermohydraulic Analysis Code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF Code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.
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Transient and Safety Analysis Code for TP-1 Sodium Cooled TWR
Volume 4: Thermal Hydraulics, 2013Co-Authors: Tenglong Cong, Wenxi Tian, Hongyang Wei, Suizheng QiuAbstract:As a new concept of reactor, traveling wave reactor (TWR) is under fundamental research. In this paper, we modeled the main parts of primary loop of TP-1 sodium cooled traveling wave reactor, which was designed by TerrraPower Co., and developed a Transient and safety Analysis Code for Sodium cooled TWR (TAST) with Fortran program. We first performed a steady Analysis based on this Code to evaluate the accuracy of this Code. The calculated results agreed well with design parameters. After this, this program was used to simulate the loss of flow accident and reactivity insertion accident separately. The transient safety Analysis indicated that TWR can work safely under these two accident conditions.Copyright © 2013 by ASME
Tenglong Cong - One of the best experts on this subject based on the ideXlab platform.
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development and preliminary validation of a steam generator 3d thermohydraulics Analysis Code staf
Nuclear Engineering and Design, 2016Co-Authors: Wenxi Tian, Tenglong Cong, Rui Zhang, G H SuAbstract:Abstract Porous media model in Fluent Code, coupled with two-phase mixture flow model, resistance model of tubes and heat transfer model through tubes, is employed to develop a steam generator thermohydraulics Analysis Code STAF (Steam generator Thermohydraulics Analysis Code based on Fluent). In this Code, the heat transfer from primary to secondary side is calculated three-dimensionally during iteration. The localized velocity, temperature, enthalpy, quality and void fraction in steam generator can be obtained by this Code. STAF is validated in two ways. First, STAF is used to calculate the thermal-hydraulic parameters in steam generator of AP 1000. The calculated results are compared with designed values to prove that the coupled heat transfer calculation in STAF is accurate. Second, STAF is employed to simulate the FRIGG test to validate the localized parameter calculation performance by comparing the calculated localized void fraction with test values.
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Analysis of westinghouse mb2 test using the steam generator thermohydraulics Analysis Code staf
Annals of Nuclear Energy, 2015Co-Authors: Tenglong Cong, Rui ZhangAbstract:Abstract In the present study, we develop a Steam-generator Thermohydraulics Analysis Code based on Fluent (STAF) for predicting the three-dimensional localized thermal–hydraulic characteristics in the primary and secondary sides of steam generator. STAF Code is developed based on the porous media model in Fluent. The flow resistances caused by the tubes, support plates, downcomer and separators are introduced to the momentum equation as additional source terms of shell side fluid; the heat transfer from primary to secondary side fluid is considered as the source term of energy equation of secondary side fluid. The flow and heat transfer in primary side, as well as the tube-to-shell-side heat transfer are solved by the user-defined functions in Fluent. STAF is used to simulate the Westinghouse MB2 test, and localized thermohydraulics parameters are obtained. The numerical results show good agreement with experimental results, demonstrating the ability of STAF to model the three-dimensional flow and heat transfer characteristics in primary and secondary side of steam generator. Besides, parameters associated with flow-induced vibration are also analyzed.
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Development and Application of a UTSG Thermal-Hydraulic Analysis Code
Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle Radioactive Waste Management and Decommissioning; Computational Fluid D, 2014Co-Authors: Tenglong Cong, Wenxi Tian, Suizheng QiuAbstract:Structural integrity of steam generator should be maintained during operation, since it performs as the pressure and heat transfer boundary of primary side coolant. Localized thermal-hydraulic parameters of secondary side are essential for the Analysis of tube wastage, fatigue and failure. In this paper, a three-dimensional thermohydraulics Analysis Code, named STAF, is developed based on FLUENT. With STAF Code, three-dimensional thermohydraulics of secondary side of AP1000 steam generator are generated. This Code is developed based on the porous media theory. In this Code, the drift flux two-phase model coupled with a simplified flow boiling model is utilized to present two-phase flow among the U-tube bundle. Downcomer, tube bundle, support plates and primary separators in steam generator are considered in STAF Code. The calculated results are compared with a general steam generator thermohydraulic Analysis Code ATHOS, which is developed by EPRI steam generator group. The comparison indicates that STAF Code performs well in evaluating thermal-hydraulic parameters in steam generator. The results show that the flow field varies significantly at different position in AP1000 steam generator. Flow vapor quality at the inlet of primary separators varies significantly, which is a severe challenge to the capacity design of separators.
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Transient and Safety Analysis Code for TP-1 Sodium Cooled TWR
Volume 4: Thermal Hydraulics, 2013Co-Authors: Tenglong Cong, Wenxi Tian, Hongyang Wei, Suizheng QiuAbstract:As a new concept of reactor, traveling wave reactor (TWR) is under fundamental research. In this paper, we modeled the main parts of primary loop of TP-1 sodium cooled traveling wave reactor, which was designed by TerrraPower Co., and developed a Transient and safety Analysis Code for Sodium cooled TWR (TAST) with Fortran program. We first performed a steady Analysis based on this Code to evaluate the accuracy of this Code. The calculated results agreed well with design parameters. After this, this program was used to simulate the loss of flow accident and reactivity insertion accident separately. The transient safety Analysis indicated that TWR can work safely under these two accident conditions.Copyright © 2013 by ASME
Goon-cherl Park - One of the best experts on this subject based on the ideXlab platform.
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Three-dimensional looped network Analysis Code including core thermal Analysis model for prismatic very high temperature gas-cooled reactor
International Journal of Thermal Sciences, 2019Co-Authors: Jeong Hun Lee, Hyoung Kyu Cho, Goon-cherl ParkAbstract:Abstract The core of a prismatic very high temperature gas-cooled reactor (VHTR) is composed of stacked graphite blocks with gaps between them, which results in undesired flows through the gaps. These flows complicate the flow distribution in the reactor core and cause difficulty in predicting the temperature distribution of the graphite block. Conventionally, computational fluid dynamics (CFD) Codes have been mainly used for the VHTR reactor core Analysis. However, they require considerable calculation time and cost, and, therefore, are considered too expensive in terms of calculation time to investigate the effect of the gap size distribution in the core. As numerous cases with different gap size combinations need to be tested in reactor design, it can be said that high calculation speed of the design Code with reasonable accuracy is an important feature. In this study, a thermo-fluid Analysis Code for the core of a prismatic VHTR, named FastNet (Flow Analysis for Steady-state Network), was developed for prediction of the core flow and temperature distribution with affordable computational cost. For rapid calculation, a flow network Analysis method was used for flow distribution Analysis, and a thermal Analysis model was added to analyze the whole core temperature distribution. To overcome the drawbacks of its low resolution, an effective thermal conductivity model and a maximum fuel temperature model were applied. Finally, to verify the Code, results of the FastNet calculation were compared to other Codes such as the CFD Code and CORONA Code as a Code-to-Code validation. The results show that a satisfactory accuracy was obtained with a remarkably short computational time.
Eric S Hendricks - One of the best experts on this subject based on the ideXlab platform.
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meanline Analysis of turbines with choked flow in the object oriented turbomachinery Analysis Code
54th AIAA Aerospace Sciences Meeting, 2016Co-Authors: Eric S HendricksAbstract:The prediction of turbomachinery performance characteristics is an important part of the conceptual aircraft engine design process. During this phase, the designer must examine the effects of a large number of turbomachinery design parameters to determine their impact on overall engine performance and weight. The lack of detailed design information available in this phase necessitates the use of simpler meanline and streamline methods to determine the turbomachinery geometry characteristics and provide performance estimates prior to more detailed CFD (Computational Fluid Dynamics) analyses. While a number of Analysis Codes have been developed for this purpose, most are written in outdated software languages and may be difficult or impossible to apply to new, unconventional designs. The Object-Oriented Turbomachinery Analysis Code (OTAC) is currently being developed at NASA Glenn Research Center to provide a flexible meanline and streamline Analysis capability in a modern object-oriented language. During the development and validation of OTAC, a limitation was identified in the Code's ability to analyze and converge turbines as the flow approached choking. This paper describes a series of changes which can be made to typical OTAC turbine meanline models to enable the assessment of choked flow up to limit load conditions. Results produced with this revised model setup are provided in the form of turbine performance maps and are compared to published maps.