Water is the most common working fluid using as the coolant for most of nuclear power plants due to its low cost, wide availability, and high heat capacity. In the Fukushima Daiichi accident caused by the tsunami following a powerful earthquake on 11th March 2011, the loss of heat sink occurred. Seawater, which had been considered as an alternative coolant before, was eventually injected into the nuclear reactor for more than one week as an emergency measure. The studies about the characteristics of heat transfer with the seawater as the working fluid have received much more attentions after the Fukushima Daiichi accident. The differences of thermophysical properties such as density, dynamic viscosity, and surface tension, between deionized water and seawater play an important role in the heat transfer. In addition, different boiling behaviors exhibit in the boiling process for two types of working fluid. Compared to pure water, the diameter of bubbles in seawater during pool boiling could be smaller than in water. To improve the safety of a nuclear reactor power plant, mechanisms of the heat transfer using seawater as an alternative emergency coolant should be studied thoroughly. In the present research, experiments of pool boiling in an annulus gap with a riser section after the sudden shutdown of the pump are investigated. Heat transfer characteristic of the heated surface using DI water and 3.5 wt% artificial seawater are described. Boiling behaviors at each fluid are compared and studied.
{"title":"Experimental Study of Boiling Characteristics of Seawater After an Accidental Shutdown of the Pump","authors":"Yuanjie Li, Shuai Ren, C. Pan","doi":"10.1115/icone2020-16648","DOIUrl":"https://doi.org/10.1115/icone2020-16648","url":null,"abstract":"\u0000 Water is the most common working fluid using as the coolant for most of nuclear power plants due to its low cost, wide availability, and high heat capacity. In the Fukushima Daiichi accident caused by the tsunami following a powerful earthquake on 11th March 2011, the loss of heat sink occurred. Seawater, which had been considered as an alternative coolant before, was eventually injected into the nuclear reactor for more than one week as an emergency measure. The studies about the characteristics of heat transfer with the seawater as the working fluid have received much more attentions after the Fukushima Daiichi accident. The differences of thermophysical properties such as density, dynamic viscosity, and surface tension, between deionized water and seawater play an important role in the heat transfer. In addition, different boiling behaviors exhibit in the boiling process for two types of working fluid. Compared to pure water, the diameter of bubbles in seawater during pool boiling could be smaller than in water. To improve the safety of a nuclear reactor power plant, mechanisms of the heat transfer using seawater as an alternative emergency coolant should be studied thoroughly. In the present research, experiments of pool boiling in an annulus gap with a riser section after the sudden shutdown of the pump are investigated. Heat transfer characteristic of the heated surface using DI water and 3.5 wt% artificial seawater are described. Boiling behaviors at each fluid are compared and studied.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122184182","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}
This paper aims to validate the effectiveness of the widely used Relap5 code in simulating two-phase natural circulation, and its capability to predict flashing-induced instabilities. The RELAP5 code is validated against experimental data from the NMR test facility, which was designed to investigate the flow instability for a BWR-type novel modular reactor (NMR). The simulations by RELAP5/MOD3.4 code had been performed under various conditions by changing system pressure, core inlet subcooling, core inlet flow resistance, and core heat power etc. The flow stability for a certain operating condition could be determined from the time trace profile of the loop natural circulation flow rate. The results showed that the simulated mass flow rate increased with increasing core inlet temperature, reproducing the experimental trend. And the maximum error between the experimental data and the calculated results is within 10%. The predicted natural circulation dimensionless numbers, the phase change number and inlet subcooling number, also had a good agreement with the experimental data. In general, the RELAP5 code is able to simulate flashing-induced instability and density wave oscillations, which occurred in the natural circulation test facility at low pressures. However, flashing tends to be suppressed at a higher pressure (400kPa). And the enlargement of core inlet resistance coefficient can also have a positive impact on natural circulation system stability.
{"title":"Validation of RELAP5/MOD3.4 for Flashing-Induced Instabilities in a Natural Circulation Loop","authors":"Yifan Xu, M. Peng, G. Xia, Yanan Zhao","doi":"10.1115/icone2020-16296","DOIUrl":"https://doi.org/10.1115/icone2020-16296","url":null,"abstract":"\u0000 This paper aims to validate the effectiveness of the widely used Relap5 code in simulating two-phase natural circulation, and its capability to predict flashing-induced instabilities. The RELAP5 code is validated against experimental data from the NMR test facility, which was designed to investigate the flow instability for a BWR-type novel modular reactor (NMR). The simulations by RELAP5/MOD3.4 code had been performed under various conditions by changing system pressure, core inlet subcooling, core inlet flow resistance, and core heat power etc. The flow stability for a certain operating condition could be determined from the time trace profile of the loop natural circulation flow rate. The results showed that the simulated mass flow rate increased with increasing core inlet temperature, reproducing the experimental trend. And the maximum error between the experimental data and the calculated results is within 10%. The predicted natural circulation dimensionless numbers, the phase change number and inlet subcooling number, also had a good agreement with the experimental data. In general, the RELAP5 code is able to simulate flashing-induced instability and density wave oscillations, which occurred in the natural circulation test facility at low pressures. However, flashing tends to be suppressed at a higher pressure (400kPa). And the enlargement of core inlet resistance coefficient can also have a positive impact on natural circulation system stability.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"93 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124636350","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}
Decontamination is a major activity in decommissioning of Nuclear Power Plants. In efforts to reduce the overall volume of nuclear waste, retrieve reusable materials, and reduce the environmental impact, many different technologies have been developed/used in prior decommissioning projects and many more are being developed. However due to the amount of technologies available and the specific use cases for each, the ability to choose an appropriate and optimal technology is a challenge. An approach was adopted to develop a tool to assist in selection of decontamination technologies appropriate for Canadian Applications. The first step is the creation of a database to compile information of the different decontamination methods currently available in one location. The next step was the development of a software program to provide a search optimization for the database based on a set of initial user conditions. The program considers a radio-isotopic breakdown of a component as identified by the user and compares its concentration (Bq/g) to regulation limits set by the Canadian Nuclear Safety Commission (CSNC) for Unconditional Clearance Levels. Then, by using the CNSC guidelines, it will determine if the component is under Unconditional Levels or not. If the component is not, the code will calculate the minimum cumulative Decontamination Factor (DFR) required to make the component compliant with unconditional requirements. The software allows for users to plan their decontamination roadmap at a present state as well as a future state where natural decay opens up the ability for a wider range of decontamination technologies and for a combination of multiple components to use a given decontamination technique.
{"title":"Development of a Selection Tool for Choosing Decontamination Technology for Canadian Applications","authors":"R. Khurmi, R. Carlisle, G. Harvel","doi":"10.1115/icone2020-16760","DOIUrl":"https://doi.org/10.1115/icone2020-16760","url":null,"abstract":"\u0000 Decontamination is a major activity in decommissioning of Nuclear Power Plants. In efforts to reduce the overall volume of nuclear waste, retrieve reusable materials, and reduce the environmental impact, many different technologies have been developed/used in prior decommissioning projects and many more are being developed. However due to the amount of technologies available and the specific use cases for each, the ability to choose an appropriate and optimal technology is a challenge. An approach was adopted to develop a tool to assist in selection of decontamination technologies appropriate for Canadian Applications. The first step is the creation of a database to compile information of the different decontamination methods currently available in one location. The next step was the development of a software program to provide a search optimization for the database based on a set of initial user conditions. The program considers a radio-isotopic breakdown of a component as identified by the user and compares its concentration (Bq/g) to regulation limits set by the Canadian Nuclear Safety Commission (CSNC) for Unconditional Clearance Levels. Then, by using the CNSC guidelines, it will determine if the component is under Unconditional Levels or not. If the component is not, the code will calculate the minimum cumulative Decontamination Factor (DFR) required to make the component compliant with unconditional requirements. The software allows for users to plan their decontamination roadmap at a present state as well as a future state where natural decay opens up the ability for a wider range of decontamination technologies and for a combination of multiple components to use a given decontamination technique.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128716718","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}
A theoretical model for Density Wave Oscillations (DWOs) flow instability in parallel rectangular channels under periodic heaving motion is established with a lumped mathematical model based on homogenous hypothesis. The parallel rectangular channels comprise of the entrance section, the heating section, the riser section and the upper- and lower plenums, which guarantee the isobaric pressure drop condition between channels and the model consists of boiling channel model, pressure drop model, parallel channel model, additional pressure drop model generated by heaving motions, the constitutive and numerical models. The effect of periodic perturbation is introduced through additional pressure drop in the momentum equation. The model is validated with experimental data of a twin-rectangular-channel flow instability experiment under static condition. Then the flow instability in parallel-rectangular-channel system is studied under periodic perturbation and the margin of flow instability and the threshold power of the system under static condition is calculated as basis condition for comparison. The effect of the amplitude and period of perturbation is analyzed analytically and the results show that the amplitude and period of perturbation shows little effect on flow instability. While when the additional pressure difference introduced by heaving motion is comparable with that under static condition, the effect of amplitude becomes stronger. And the period of perturbation strongly effects the threshold power when it is identical to that of natural period of the system, which can be explained by resonance between the perturbation and the system. And this effect is even stronger when the asymmetric heating condition is introduced.
{"title":"Theoretical Research on Two-Phase Flow Instability in Parallel Rectangular Channels Under Periodic Perturbation","authors":"Libo Qian, Jian Deng, Tao Huang, R. Cai","doi":"10.1115/icone2020-16766","DOIUrl":"https://doi.org/10.1115/icone2020-16766","url":null,"abstract":"\u0000 A theoretical model for Density Wave Oscillations (DWOs) flow instability in parallel rectangular channels under periodic heaving motion is established with a lumped mathematical model based on homogenous hypothesis. The parallel rectangular channels comprise of the entrance section, the heating section, the riser section and the upper- and lower plenums, which guarantee the isobaric pressure drop condition between channels and the model consists of boiling channel model, pressure drop model, parallel channel model, additional pressure drop model generated by heaving motions, the constitutive and numerical models. The effect of periodic perturbation is introduced through additional pressure drop in the momentum equation. The model is validated with experimental data of a twin-rectangular-channel flow instability experiment under static condition. Then the flow instability in parallel-rectangular-channel system is studied under periodic perturbation and the margin of flow instability and the threshold power of the system under static condition is calculated as basis condition for comparison. The effect of the amplitude and period of perturbation is analyzed analytically and the results show that the amplitude and period of perturbation shows little effect on flow instability. While when the additional pressure difference introduced by heaving motion is comparable with that under static condition, the effect of amplitude becomes stronger. And the period of perturbation strongly effects the threshold power when it is identical to that of natural period of the system, which can be explained by resonance between the perturbation and the system. And this effect is even stronger when the asymmetric heating condition is introduced.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"46 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127580633","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}
K. Ono, Yasunori Yamamoto, Masayoshi Mori, Tetsuya Takada
Isolation condensers (ICs) are important passive cooling systems in BWRs. After the Fukushima Daiichi Nuclear Power Station accident, concerns if the IC was able to restart with the inflow of hydrogen were arose. Because ICs lose heat removal ability when non-condensable gas inflow occurs, accurate evaluation of the effect is necessary. To develop analysis methods, as an initial stage, experiments and analyses considering only high-pressure steam and water were conducted. The experiment was done by an isolation condenser simulator which contains an accumulator with heaters inside, and a heat transfer tube. From the experiment, all steam was condensed at the heat transfer tube and the approximate position of complete condensation was confirmed from the temperature distribution and the observation. The experiment provided data such as temperature distribution, natural circulation flow rate, and pressure to compare with the analysis. The analyses were conducted for 4 cases of void fraction values at the heat transfer tube inlet and found that it has a high sensitivity to condensation. The reason is estimated to be the difference in inflow velocity that strongly depends on the void fraction even if the mass flow rate is constant. And the initial condition of the liquid film also affected condensation process. Heat removal at the section before the heat transfer tube should be considered to adjust void fraction at the inlet of heat transfer tube.
{"title":"Experiment and Analysis on Isolation Condenser Simulator Using Pressurized Steam","authors":"K. Ono, Yasunori Yamamoto, Masayoshi Mori, Tetsuya Takada","doi":"10.1115/icone2020-16842","DOIUrl":"https://doi.org/10.1115/icone2020-16842","url":null,"abstract":"\u0000 Isolation condensers (ICs) are important passive cooling systems in BWRs. After the Fukushima Daiichi Nuclear Power Station accident, concerns if the IC was able to restart with the inflow of hydrogen were arose. Because ICs lose heat removal ability when non-condensable gas inflow occurs, accurate evaluation of the effect is necessary. To develop analysis methods, as an initial stage, experiments and analyses considering only high-pressure steam and water were conducted. The experiment was done by an isolation condenser simulator which contains an accumulator with heaters inside, and a heat transfer tube. From the experiment, all steam was condensed at the heat transfer tube and the approximate position of complete condensation was confirmed from the temperature distribution and the observation. The experiment provided data such as temperature distribution, natural circulation flow rate, and pressure to compare with the analysis. The analyses were conducted for 4 cases of void fraction values at the heat transfer tube inlet and found that it has a high sensitivity to condensation. The reason is estimated to be the difference in inflow velocity that strongly depends on the void fraction even if the mass flow rate is constant. And the initial condition of the liquid film also affected condensation process. Heat removal at the section before the heat transfer tube should be considered to adjust void fraction at the inlet of heat transfer tube.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"32 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130730259","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}
Many of the analytical codes used in the nuclear industry, such as TRACE, RELAP5, and PARCS, approximate the equations that model the physics via a linearized system of equations. One common difficulty when solving linearized systems is that an accurately formulated system of equations may be ill-conditioned. Ill-conditioned matrices can result in significant amplification of error leading to poor, or even invalid, results. Ill-conditioned matrices lead to some challenging issues for the analytical code developers: • An ill-conditioned matrix is often solvable, and there may be no obvious indication numerically that something has gone wrong even though numerical error is large. Thus, how can ill-conditioning be effectively detected for a matrix? • When ill-conditioning is detected, how can the source of the ill-conditioning be determined so that it can be analyzed and corrected? Ill-conditioning is fundamentally a geometric problem that can be understood with geometric concepts associated with matrices and vectors. Geometric concepts and tools, useful for understanding the cause of ill-conditioning of a matrix, are presented. A geometric understanding of ill-conditioning can point to the rows or columns of the matrix that most contribute to ill-conditioning so that the source of ill-conditioning can be analyzed and understood, and leads to techniques for building matrix preconditioners to improve the solvability of the matrix.
{"title":"Identifying the Cause of and Fixing Ill-Conditioned Matrices in Nuclear Analysis Codes","authors":"Lance C. Larsen","doi":"10.1115/icone2020-16903","DOIUrl":"https://doi.org/10.1115/icone2020-16903","url":null,"abstract":"\u0000 Many of the analytical codes used in the nuclear industry, such as TRACE, RELAP5, and PARCS, approximate the equations that model the physics via a linearized system of equations. One common difficulty when solving linearized systems is that an accurately formulated system of equations may be ill-conditioned. Ill-conditioned matrices can result in significant amplification of error leading to poor, or even invalid, results. Ill-conditioned matrices lead to some challenging issues for the analytical code developers:\u0000 • An ill-conditioned matrix is often solvable, and there may be no obvious indication numerically that something has gone wrong even though numerical error is large. Thus, how can ill-conditioning be effectively detected for a matrix?\u0000 • When ill-conditioning is detected, how can the source of the ill-conditioning be determined so that it can be analyzed and corrected?\u0000 Ill-conditioning is fundamentally a geometric problem that can be understood with geometric concepts associated with matrices and vectors. Geometric concepts and tools, useful for understanding the cause of ill-conditioning of a matrix, are presented. A geometric understanding of ill-conditioning can point to the rows or columns of the matrix that most contribute to ill-conditioning so that the source of ill-conditioning can be analyzed and understood, and leads to techniques for building matrix preconditioners to improve the solvability of the matrix.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134106287","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}
Irradiation hardening is one of the most important aging effects of reactor pressure vessel (RPV) steel in long-term service. A number of studies have indicated that irradiation hardening is mainly caused by irradiation induced defects, such as dislocation loops and precipitates. In this paper, we have simulated the irradiation damage of low-copper reactor pressure vessel (RPV) steel. The generation of Mn-Ni-Si precipitates is simulated by the cluster dynamics model based on rate theory. On this basis, the crystal plasticity finite element method based on modified crystal plasticity model is used to simulate the effect of Mn-Ni-Si precipitates on the mechanical properties of RPV steels. The simulated results has been compared with the experimental results from the literature. By coupling the cluster dynamics and the crystal plastic finite element method, we suggest a multi-scale simulation method to simulate and predict irradiation damage of RPV steel.
{"title":"Study on Irradiation Damage of RPV Steels Based on Coupling Cluster Dynamics and Crystal Plasticity Finite Element Method","authors":"Xiaotong Wang, Ying Luo, Yu-Qin Dong, W. Yao","doi":"10.1115/icone2020-16272","DOIUrl":"https://doi.org/10.1115/icone2020-16272","url":null,"abstract":"\u0000 Irradiation hardening is one of the most important aging effects of reactor pressure vessel (RPV) steel in long-term service. A number of studies have indicated that irradiation hardening is mainly caused by irradiation induced defects, such as dislocation loops and precipitates. In this paper, we have simulated the irradiation damage of low-copper reactor pressure vessel (RPV) steel. The generation of Mn-Ni-Si precipitates is simulated by the cluster dynamics model based on rate theory. On this basis, the crystal plasticity finite element method based on modified crystal plasticity model is used to simulate the effect of Mn-Ni-Si precipitates on the mechanical properties of RPV steels. The simulated results has been compared with the experimental results from the literature. By coupling the cluster dynamics and the crystal plastic finite element method, we suggest a multi-scale simulation method to simulate and predict irradiation damage of RPV steel.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"145 11","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132124139","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}
Fang Hongyu, L. Wang, Zhu Dahuan, C. Rong, Jiang Xiaowei, Zhang Dan, Libo Qian, Dan Wu
With the development of the nuclear energy industry, small modular reactors (SMRs) have become an important option in China’s energy development due to their advantages in terms of safety and economics. The helical-coil steam generator is an important part of SMR, and the structure parameters need to be optimized urgently to reduce costs and improve safety. In this paper, various thermohydraulic parameters were used as indicators for the design quality of the helical-coil steam generator, such as the volume of the equipment, velocity of the flow, maximum temperature of the tubes, and natural circulation capability. The optimization of these target parameters has important theoretical and practical values. The genetic algorithm method is widely used to processes data efficiently, and it is able to reach a optimum design rapidly, which makes it a good choice for the optimization of parameters. The results show that this method could efficiently reduce the volume of the helical coil steam generator.
{"title":"Optimization Design for the Configuration Parameters of Helical-Coil Steam Generator Volume","authors":"Fang Hongyu, L. Wang, Zhu Dahuan, C. Rong, Jiang Xiaowei, Zhang Dan, Libo Qian, Dan Wu","doi":"10.1115/icone2020-16891","DOIUrl":"https://doi.org/10.1115/icone2020-16891","url":null,"abstract":"\u0000 With the development of the nuclear energy industry, small modular reactors (SMRs) have become an important option in China’s energy development due to their advantages in terms of safety and economics. The helical-coil steam generator is an important part of SMR, and the structure parameters need to be optimized urgently to reduce costs and improve safety.\u0000 In this paper, various thermohydraulic parameters were used as indicators for the design quality of the helical-coil steam generator, such as the volume of the equipment, velocity of the flow, maximum temperature of the tubes, and natural circulation capability. The optimization of these target parameters has important theoretical and practical values. The genetic algorithm method is widely used to processes data efficiently, and it is able to reach a optimum design rapidly, which makes it a good choice for the optimization of parameters. The results show that this method could efficiently reduce the volume of the helical coil steam generator.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124089706","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}
Tight lattice bundle can improve the conversion ratio and the heat transfer coefficient between the fuel bundle and the coolant, which is widely used in the innovative reactor fuel bundle design. The P/D ratio of a tight lattice bundle is usually less than 1.1, which is smaller than that of a conventional rod bundle. In the small-break loss-of-coolant accident (LOCA), the steam-water two-phase flow will occur in the reactor. The investigation of gas-liquid two-phase flow in the tight lattice is very important to the reactor safety analysis. A dual sub-channels tight lattice was designed in this study. The original reference of the channel is the annular fuel bundle, with the fuel diameter of 15.52mm, pitch of 16.51mm, P/D = 1.06. The original reference of working condition is the stream-water two-phase flow under the pressure of 15.5MPa. The experimental condition is the air-water two-phase flow at the normal temperature and pressure. According to the ratio of a critical bubble diameter in the reactor (steam-water) to that in atmospheric conditions (air-water), the channel is zoomed in 2.7 times. The diameter of the rod in the dual sub-channels tight lattice is 42mm and the pitch is 44.6mm. The total length of the dual sub-channels tight lattice is 3m. A self-developed 16 × 32 Wire-mesh sensor (WMS) was used to measure the void fraction distribution of air-water two-phase flow in the dual sub-channels tight lattice channel. The spatial resolution of the WMS is 2.79mm and the temporal resolution is 5000fps. The WMS was installed at a distance of 2.5m from the channel inlet and 0.5m from the outlet, which can avoid the influence of outlet on bubbles. The experimental range of flow condition is 0.921–1.84m/s for the superficial liquid velocity and 0.0884–1.07m/s for the superficial gas velocity. The instantaneous and time-averaged void fraction distributions in the channel was measured. With the increase of superficial gas velocity, the distribution of void fraction distribution changed from the wall peak to the core peak. The characteristics of bubbles in the sub-channel were also discussed in this study.
{"title":"An Experimental Study of Two-Phase Flow in a Tight Lattice Using Wire-Mesh Sensor","authors":"Hengwei Zhang, Yao Xiao, H. Gu","doi":"10.1115/icone2020-16211","DOIUrl":"https://doi.org/10.1115/icone2020-16211","url":null,"abstract":"\u0000 Tight lattice bundle can improve the conversion ratio and the heat transfer coefficient between the fuel bundle and the coolant, which is widely used in the innovative reactor fuel bundle design. The P/D ratio of a tight lattice bundle is usually less than 1.1, which is smaller than that of a conventional rod bundle. In the small-break loss-of-coolant accident (LOCA), the steam-water two-phase flow will occur in the reactor. The investigation of gas-liquid two-phase flow in the tight lattice is very important to the reactor safety analysis. A dual sub-channels tight lattice was designed in this study. The original reference of the channel is the annular fuel bundle, with the fuel diameter of 15.52mm, pitch of 16.51mm, P/D = 1.06. The original reference of working condition is the stream-water two-phase flow under the pressure of 15.5MPa. The experimental condition is the air-water two-phase flow at the normal temperature and pressure. According to the ratio of a critical bubble diameter in the reactor (steam-water) to that in atmospheric conditions (air-water), the channel is zoomed in 2.7 times. The diameter of the rod in the dual sub-channels tight lattice is 42mm and the pitch is 44.6mm. The total length of the dual sub-channels tight lattice is 3m. A self-developed 16 × 32 Wire-mesh sensor (WMS) was used to measure the void fraction distribution of air-water two-phase flow in the dual sub-channels tight lattice channel. The spatial resolution of the WMS is 2.79mm and the temporal resolution is 5000fps. The WMS was installed at a distance of 2.5m from the channel inlet and 0.5m from the outlet, which can avoid the influence of outlet on bubbles. The experimental range of flow condition is 0.921–1.84m/s for the superficial liquid velocity and 0.0884–1.07m/s for the superficial gas velocity. The instantaneous and time-averaged void fraction distributions in the channel was measured. With the increase of superficial gas velocity, the distribution of void fraction distribution changed from the wall peak to the core peak. The characteristics of bubbles in the sub-channel were also discussed in this study.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114839290","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}
Yikai Wu, Wenxuan Ju, Yusheng Liu, F. Zhao, Sichao Tan
The single droplet phase change model during motion is developed based on the phenomena description and mechanism comprehension, which including the droplet phase change model as well as the droplet motion model. Then, the calculation of the droplet phase change characteristics during moving in the uniform flow in the gravity separation space is conducted. The results show that when the droplet are evaporating during its moving, the radius will decrease continuously and it will be carried more easily by the steam vapor, which will lead to the larger separation radii of the droplets and the reduced the gravity separation efficiency. In addition, this paper shows the three-dimensional map for the critical separation over the pressure difference and the steam vapor flow velocity, which can contribute to forecast the influence of the droplet phase change on the separation characteristics. The results can be applied in the design of the steam-water separation plants.
{"title":"Critical Conditions for Secondary Droplets Generated by Droplets Colliding Walls With Different Angles","authors":"Yikai Wu, Wenxuan Ju, Yusheng Liu, F. Zhao, Sichao Tan","doi":"10.1115/icone2020-16864","DOIUrl":"https://doi.org/10.1115/icone2020-16864","url":null,"abstract":"\u0000 The single droplet phase change model during motion is developed based on the phenomena description and mechanism comprehension, which including the droplet phase change model as well as the droplet motion model. Then, the calculation of the droplet phase change characteristics during moving in the uniform flow in the gravity separation space is conducted. The results show that when the droplet are evaporating during its moving, the radius will decrease continuously and it will be carried more easily by the steam vapor, which will lead to the larger separation radii of the droplets and the reduced the gravity separation efficiency. In addition, this paper shows the three-dimensional map for the critical separation over the pressure difference and the steam vapor flow velocity, which can contribute to forecast the influence of the droplet phase change on the separation characteristics. The results can be applied in the design of the steam-water separation plants.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"44 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125828099","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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