Impinging jets are commonly utilized in the run-out table (ROT) cooling in the hot rolling process in steel manufacturing industries. The phenomenon of rapid cooling of a sufficiently hot surface is termed as the quenching. The present paper reports the rewetting behavior of 0.15 mm thick hot moving stainless steel foil (SS-304) by circular impinging jet from bottom side through experimental investigation. The transient temperature of the hot foil is recorded by using thermal imaging camera (A655sc, FLIR system). Tests are performed for a varied range of Reynolds number (Re = 2500–10000), nozzle to plate distance (z/d = 6), moving plate velocity (0–40 mm/s) and initial surface temperature 500±10 °C. Transient temperature obtained from thermal imaging camera is used to evaluate rewetting time and rewetting velocity. Based on the experimental investigation correlation has been proposed to predict non-dimensional rewetting velocity as a function of various parameters, namely, Reynolds number, non-dimensional axial distance and moving plate velocity.
{"title":"Rewetting Analysis of Hot Moving Surface by Round Water Jet Impingement","authors":"A. Sharma, Mayank Modak, S. Sahu","doi":"10.1115/ICONE26-81673","DOIUrl":"https://doi.org/10.1115/ICONE26-81673","url":null,"abstract":"Impinging jets are commonly utilized in the run-out table (ROT) cooling in the hot rolling process in steel manufacturing industries. The phenomenon of rapid cooling of a sufficiently hot surface is termed as the quenching. The present paper reports the rewetting behavior of 0.15 mm thick hot moving stainless steel foil (SS-304) by circular impinging jet from bottom side through experimental investigation. The transient temperature of the hot foil is recorded by using thermal imaging camera (A655sc, FLIR system). Tests are performed for a varied range of Reynolds number (Re = 2500–10000), nozzle to plate distance (z/d = 6), moving plate velocity (0–40 mm/s) and initial surface temperature 500±10 °C. Transient temperature obtained from thermal imaging camera is used to evaluate rewetting time and rewetting velocity. Based on the experimental investigation correlation has been proposed to predict non-dimensional rewetting velocity as a function of various parameters, namely, Reynolds number, non-dimensional axial distance and moving plate velocity.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130095827","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Decommissioning of nuclear systems and consolidation of resulting waste remains a major hurdle for the industry. The radioactive nature of nuclear waste makes manual handling unfeasible, while teleoperation and inflexible automation also have drawbacks. This effort utilizes virtual fixtures to augment path creation for non-contact material reduction tasks. It builds on previous Nuclear and Applied Robotics Group work with Variable Normal Surface Virtual Fixtures, which generate surfaces offset from task surfaces. Offset surfaces can then provide poses at a set orientation to task surfaces. Once the Variable Normal Surface Virtual Fixture is generated from task information, operators build a list of desired path poses. The Robot Operating System Descartes package takes the pose list and plans a smooth trajectory for task execution. Planar and cylindrical demonstrations based on experimental studies at the United Kingdom’s Sellafield site were performed. This methodology augments waste reduction by allowing flexible laser cutting routes.
{"title":"Virtual Fixture Augmentation of Operator Selection of Non-Contact Material Reduction Task Paths","authors":"Andrew Sharp, Christina Petlowany, M. Pryor","doi":"10.1115/ICONE26-82398","DOIUrl":"https://doi.org/10.1115/ICONE26-82398","url":null,"abstract":"Decommissioning of nuclear systems and consolidation of resulting waste remains a major hurdle for the industry. The radioactive nature of nuclear waste makes manual handling unfeasible, while teleoperation and inflexible automation also have drawbacks.\u0000 This effort utilizes virtual fixtures to augment path creation for non-contact material reduction tasks. It builds on previous Nuclear and Applied Robotics Group work with Variable Normal Surface Virtual Fixtures, which generate surfaces offset from task surfaces. Offset surfaces can then provide poses at a set orientation to task surfaces. Once the Variable Normal Surface Virtual Fixture is generated from task information, operators build a list of desired path poses. The Robot Operating System Descartes package takes the pose list and plans a smooth trajectory for task execution. Planar and cylindrical demonstrations based on experimental studies at the United Kingdom’s Sellafield site were performed. This methodology augments waste reduction by allowing flexible laser cutting routes.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133662242","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation. The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated. The aim was to identify a group structure with very low dependency on the actual reactor spectrum.
{"title":"Analysis of Major Group Structures Used for Nuclear Reactor Simulations","authors":"M. D. Filippo, J. Křepel, K. Mikityuk, H. Prasser","doi":"10.1115/ICONE26-81445","DOIUrl":"https://doi.org/10.1115/ICONE26-81445","url":null,"abstract":"Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation.\u0000 The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated.\u0000 The aim was to identify a group structure with very low dependency on the actual reactor spectrum.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"191 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122353227","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
R. Nistor-Vlad, D. Dupleac, I. Prisecaru, C. Allison, M. Pérez-Ferragut, J. Hohorst
RELAP/SCDAPSIM is a best-estimate nuclear tool designed to analyze the behaviour of reactor systems during normal and accident conditions. Three main versions of RELAP/SCDAPSIM are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.4 is the version of the code used by licensed users and program members for critical applications such as research reactors and nuclear power plant applications. Even though the code was initially designed for LWRs (Light Water Reactors), Politehnica University of Bucharest demonstrated the applicability of the RELAP/SCDAPSIM code for CANDU (CANada Deuterium Uranium) reactors analyses, by simulating some of the most important postulated accident transients (i.e. large break LOCA, main steam line break, the natural circulation in the heat transport system). Current trends refer to the BEPU (Best Estimate Plus Uncertainty) approach in the safety analysis of nuclear reactors. BEPU is a modern and technically consistent approach has been built upon best estimate methods including an evaluation of the uncertainty in the calculated results. ISS and UPC started the development of an uncertainty evaluation package to RELAP/SCDAPSIM/MOD4.0 code version which is currently implemented in MOD3.4 version of the code also. The uncertainty evaluation capability is implemented as an alternative run mode, the “uncertainty” mode, which allows the automatic execution of an uncertainty analysis based on the probabilistic approach. A complete uncertainty analysis using RELAP/SCDAPSIM/MOD3.4 code requires the execution of three related phases, namely the “setup” phase, the “simulation” phase consisting of several executions, and the “postprocessing” phase. The uncertainty data has to be supplied for the two types of parameters, the “input treatable” and the “source correlation” quantities. The required information is the probability distribution function and its characteristic parameters. This paper is mainly focused on the application of the uncertainty package in CANDU reactors accident analysis and it describes the steps to perform the uncertainty analysis, the uncertainty selected parameters (input treatable and source correlation parameters), the calculation results with RELAP/SCDAPSIM and some conclusions.
{"title":"CANDU 6 Accident Analysis Using RELAP/SCDAPSIM With the Integrated Uncertainty Package","authors":"R. Nistor-Vlad, D. Dupleac, I. Prisecaru, C. Allison, M. Pérez-Ferragut, J. Hohorst","doi":"10.1115/ICONE26-82241","DOIUrl":"https://doi.org/10.1115/ICONE26-82241","url":null,"abstract":"RELAP/SCDAPSIM is a best-estimate nuclear tool designed to analyze the behaviour of reactor systems during normal and accident conditions. Three main versions of RELAP/SCDAPSIM are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.4 is the version of the code used by licensed users and program members for critical applications such as research reactors and nuclear power plant applications. Even though the code was initially designed for LWRs (Light Water Reactors), Politehnica University of Bucharest demonstrated the applicability of the RELAP/SCDAPSIM code for CANDU (CANada Deuterium Uranium) reactors analyses, by simulating some of the most important postulated accident transients (i.e. large break LOCA, main steam line break, the natural circulation in the heat transport system). Current trends refer to the BEPU (Best Estimate Plus Uncertainty) approach in the safety analysis of nuclear reactors. BEPU is a modern and technically consistent approach has been built upon best estimate methods including an evaluation of the uncertainty in the calculated results. ISS and UPC started the development of an uncertainty evaluation package to RELAP/SCDAPSIM/MOD4.0 code version which is currently implemented in MOD3.4 version of the code also. The uncertainty evaluation capability is implemented as an alternative run mode, the “uncertainty” mode, which allows the automatic execution of an uncertainty analysis based on the probabilistic approach. A complete uncertainty analysis using RELAP/SCDAPSIM/MOD3.4 code requires the execution of three related phases, namely the “setup” phase, the “simulation” phase consisting of several executions, and the “postprocessing” phase. The uncertainty data has to be supplied for the two types of parameters, the “input treatable” and the “source correlation” quantities. The required information is the probability distribution function and its characteristic parameters. This paper is mainly focused on the application of the uncertainty package in CANDU reactors accident analysis and it describes the steps to perform the uncertainty analysis, the uncertainty selected parameters (input treatable and source correlation parameters), the calculation results with RELAP/SCDAPSIM and some conclusions.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125282248","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The event at the Fukushima Daiichi Spent Fuel Pools (SFPs) has renewed interest in quantifying the safety margins related to loss of coolant accidents in Irradiated Fuel Bays (IFBs). Thermal-hydraulic analyses of exposed spent CANDU fuel has been limited to a small number of bundles due to its complex bundle geometry and open rack design. This paper presents a process to predict the steady state temperature and velocity of air as it passes through a rack of spent fuel using analytical models and Computational Fluid Dynamics (CFDs) techniques. The scenario acts as lower bound estimate for the effectiveness of convection during a complete loss of coolant in a fuel bay by examining the heat-up of a stand-alone rack without flow resistance of the bundles. The correct incorporation of flow resistance is a necessary step before conclusions are made about the available safety margins of irradiated fuel bays.
{"title":"Convective Heat Transfer in CANDU Spent Fuel Racks After a Loss of Coolant","authors":"Derek Logtenberg, W. Grant, P. Chan, E. Corcoran","doi":"10.1115/ICONE26-81461","DOIUrl":"https://doi.org/10.1115/ICONE26-81461","url":null,"abstract":"The event at the Fukushima Daiichi Spent Fuel Pools (SFPs) has renewed interest in quantifying the safety margins related to loss of coolant accidents in Irradiated Fuel Bays (IFBs). Thermal-hydraulic analyses of exposed spent CANDU fuel has been limited to a small number of bundles due to its complex bundle geometry and open rack design. This paper presents a process to predict the steady state temperature and velocity of air as it passes through a rack of spent fuel using analytical models and Computational Fluid Dynamics (CFDs) techniques. The scenario acts as lower bound estimate for the effectiveness of convection during a complete loss of coolant in a fuel bay by examining the heat-up of a stand-alone rack without flow resistance of the bundles. The correct incorporation of flow resistance is a necessary step before conclusions are made about the available safety margins of irradiated fuel bays.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125596094","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Xiaohan Zhao, Mingjun Wang, W. Tian, G. Su, S. Qiu
Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It’s meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.
{"title":"Hydraulic Characteristics Research on SG Under Tube Plugging Operations Using FLUENT","authors":"Xiaohan Zhao, Mingjun Wang, W. Tian, G. Su, S. Qiu","doi":"10.1115/ICONE26-81641","DOIUrl":"https://doi.org/10.1115/ICONE26-81641","url":null,"abstract":"Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It’s meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"32 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114155558","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Li Xing, Sichao Tan, Zhengpeng Mi, Peiyao Qi, Huang Yunlong
Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.
{"title":"Flow and Temperature Fields Measurement Inside Rod Bundle by the Combined Use of PIV and LIF Technique","authors":"Li Xing, Sichao Tan, Zhengpeng Mi, Peiyao Qi, Huang Yunlong","doi":"10.1115/ICONE26-81526","DOIUrl":"https://doi.org/10.1115/ICONE26-81526","url":null,"abstract":"Thermal hydraulic research of reactor core is important in nuclear energy applications, the flow and heat transfer characteristics of coolant in reactor fuel assembly has a great influence on the performance and safety of nuclear power plants. Particle image velocimetry (PIV) and Laser induced fluorescence (LIF) are the instantaneous, non-intrusive, whole-field fluid mechanics measuring method. In this study, the simultaneous measurement of flow field and temperature field for a rod bundle was conducted using PIV and LIF technique. A facility system, utilizing the matching index of refraction approach, has been designed and constructed for the measurement of velocity and temperature in the rod bundle. In order for further study on complex heat and mass transfer characteristic of rod bundle, the single-phase experiments on the heating conditions are performed. One of unique characteristics of the velocity and temperature distribution downstream the spacer grid was obtained. The experimental results show that the combined use of PIV and LIF technique is applied to the measurement of multi-physical field in a rod bundle is feasible, the measuring characteristics of non-intrusive ensured accuracy of whole field data. The whole field experimental data obtained in rod bundle benefits the design of spacer grid geometry.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"27 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115322342","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The study on the deposition of graphite dust is significant to the safety of High-Temperature Gas-cooled Reactor (HTGR) due to potential accident such as localized hot-spots and intensity change which is caused by the graphite dust generated by abrasion of fuel elements. Based on the steady flow and three-dimensional face centered structures of fuel pebble bed, the discrete phase model (DPM) were applied to simulate trajectory of graphite dust in conditions of HTGR. To determinate the deposition of particle, the present study introduces a rebound condition with critical velocity by a user defined function. The particle trajectories show most of particle deposition can be summed up as the effect of backflow region, turbulent diffusion and inertial impact. The original trap condition overestimates the deposition fraction especially for large particles compared with involving rebound condition. In addition, the trend of deposition fraction shows as the dimeter of particle increases, deposition fraction decreases first and then increases.
{"title":"A Numerical Study of Particle Deposition Through Fuel Pebble Bed in HTGR","authors":"Qi Sun, Gang Zhao, W. Peng, Suyuan Yu","doi":"10.1115/ICONE26-81792","DOIUrl":"https://doi.org/10.1115/ICONE26-81792","url":null,"abstract":"The study on the deposition of graphite dust is significant to the safety of High-Temperature Gas-cooled Reactor (HTGR) due to potential accident such as localized hot-spots and intensity change which is caused by the graphite dust generated by abrasion of fuel elements. Based on the steady flow and three-dimensional face centered structures of fuel pebble bed, the discrete phase model (DPM) were applied to simulate trajectory of graphite dust in conditions of HTGR. To determinate the deposition of particle, the present study introduces a rebound condition with critical velocity by a user defined function. The particle trajectories show most of particle deposition can be summed up as the effect of backflow region, turbulent diffusion and inertial impact. The original trap condition overestimates the deposition fraction especially for large particles compared with involving rebound condition. In addition, the trend of deposition fraction shows as the dimeter of particle increases, deposition fraction decreases first and then increases.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116056845","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Analysis of engineering approach to the operational life forecasting for constructional elements with respect to the low-cycle fatigue is carried out. Applicability limits for a hypothesis on existence of generalized cyclic-deforming diagram in case of complex low-cycle loading (deforming) are shown. It is determined, that under condition of plane-stress state and piecewise-broken trajectories of cycle loading with stresses and deformation checking the cyclic deforming diagram is united in limits of deformations, which are not exceeded 10 values of deformation corresponding material yield point. Generalized kinematic equation of material damageability is described. The method of damageability parameter utilization for increasing of accuracy calculation of structural elements low-cycle fatigue by using the effective coefficients of stresses and deformations taking into account the damageability parameter is given.
{"title":"Low-Cycle Strength of Elements of Constructions","authors":"A. Zvorykin, Roman Popov, M. Bobyr, I. Pioro","doi":"10.1115/ICONE26-81860","DOIUrl":"https://doi.org/10.1115/ICONE26-81860","url":null,"abstract":"Analysis of engineering approach to the operational life forecasting for constructional elements with respect to the low-cycle fatigue is carried out. Applicability limits for a hypothesis on existence of generalized cyclic-deforming diagram in case of complex low-cycle loading (deforming) are shown.\u0000 It is determined, that under condition of plane-stress state and piecewise-broken trajectories of cycle loading with stresses and deformation checking the cyclic deforming diagram is united in limits of deformations, which are not exceeded 10 values of deformation corresponding material yield point. Generalized kinematic equation of material damageability is described.\u0000 The method of damageability parameter utilization for increasing of accuracy calculation of structural elements low-cycle fatigue by using the effective coefficients of stresses and deformations taking into account the damageability parameter is given.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131207784","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
E. Osigwe, A. Gad-Briggs, T. Nikolaidis, P. Pilidis, S. Sampath
One major challenge to the accurate development of performance simulation tool for component-based nuclear power plant engine models is the difficulty in accessing component performance maps; hence, researchers or engineers often rely on estimation approach using various scaling techniques. This paper describes a multi-fluid scaling approach used to determine the component characteristics of a closed-cycle gas turbine plant from an existing component map with their design data, which can be applied for different working fluids as may be required in closed-cycle gas turbine operations to adapt data from one component map into a new component map. Each component operation is defined by an appropriate change of state equations which describes its thermodynamic behavior, thus, a consideration of the working fluid properties is of high relevance to the scaling approach. The multi-fluid scaling technique described in this paper was used to develop a computer simulation tool called GT-ACYSS, which can be valuable for analyzing the performance of closed-cycle gas turbine operations with different working fluids. This approach makes it easy to theoretically scale existing map using similar or different working fluids without carrying out a full experimental test or repeating the whole design and development process. The results of selected case studies show a reasonable agreement with available data.
{"title":"Multi-Fluid Gas Turbine Components Scaling for a Generation IV Nuclear Power Plant Performance Simulation","authors":"E. Osigwe, A. Gad-Briggs, T. Nikolaidis, P. Pilidis, S. Sampath","doi":"10.1115/ICONE26-82373","DOIUrl":"https://doi.org/10.1115/ICONE26-82373","url":null,"abstract":"One major challenge to the accurate development of performance simulation tool for component-based nuclear power plant engine models is the difficulty in accessing component performance maps; hence, researchers or engineers often rely on estimation approach using various scaling techniques. This paper describes a multi-fluid scaling approach used to determine the component characteristics of a closed-cycle gas turbine plant from an existing component map with their design data, which can be applied for different working fluids as may be required in closed-cycle gas turbine operations to adapt data from one component map into a new component map. Each component operation is defined by an appropriate change of state equations which describes its thermodynamic behavior, thus, a consideration of the working fluid properties is of high relevance to the scaling approach. The multi-fluid scaling technique described in this paper was used to develop a computer simulation tool called GT-ACYSS, which can be valuable for analyzing the performance of closed-cycle gas turbine operations with different working fluids. This approach makes it easy to theoretically scale existing map using similar or different working fluids without carrying out a full experimental test or repeating the whole design and development process. The results of selected case studies show a reasonable agreement with available data.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"159 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133530937","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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