Heat transfer limit is one of the main concerns of IVR-ERVC strategy. When the liquid level in the system is so low that natural circulation cannot be formed, the coolant near the outer surface of the reactor pressure vessel lower head is in the pool boiling state. In this research a one-dimensional full-height experimental facility was established to research the heat transfer limit (CHF) of ERVC under the pool boiling condition with a one-dimensional heating block, which is used to simulate the lower head of reactor pressure vessel. The experiment was carried out at different liquid levels and the results are compared with those of natural circulation experiment at the same liquid level. Experimental results show that CHF increases with the increase of the inclination angle of heating block. Meanwhile, the increase of liquid level is beneficial to the improvement of CHF. In addition, it can be found that the influence of flow path size on CHF is complex, and CHF does not change monotonically with the increase of flow path size. On the other hand, compared with the results of natural circulation at the same liquid level, the CHF values under pool boiling are relatively low. These results are expected to improve the understanding of IVR-ERVC strategy.
{"title":"Critical Heat Flux Experiments for IVR-ERVC Strategy Under the Pool Boiling Condition","authors":"G. Wang, B. Kuang, Yihai He, Pengfei Liu","doi":"10.1115/icone29-93623","DOIUrl":"https://doi.org/10.1115/icone29-93623","url":null,"abstract":"\u0000 Heat transfer limit is one of the main concerns of IVR-ERVC strategy. When the liquid level in the system is so low that natural circulation cannot be formed, the coolant near the outer surface of the reactor pressure vessel lower head is in the pool boiling state. In this research a one-dimensional full-height experimental facility was established to research the heat transfer limit (CHF) of ERVC under the pool boiling condition with a one-dimensional heating block, which is used to simulate the lower head of reactor pressure vessel. The experiment was carried out at different liquid levels and the results are compared with those of natural circulation experiment at the same liquid level. Experimental results show that CHF increases with the increase of the inclination angle of heating block. Meanwhile, the increase of liquid level is beneficial to the improvement of CHF. In addition, it can be found that the influence of flow path size on CHF is complex, and CHF does not change monotonically with the increase of flow path size. On the other hand, compared with the results of natural circulation at the same liquid level, the CHF values under pool boiling are relatively low. These results are expected to improve the understanding of IVR-ERVC strategy.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"50 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132588417","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 studies the structural parameters of fuel assemblies suitable for high neutron flux density environments. Due to the high neutron flux density in the core, high-flux reactors provide an experimental environment for neutron irradiation. However, high neutron flux leads to high heat flux density on the fuel assembly surface, which brings challenges to the design of fuel assembly. Therefore, it is very important to study the structural design of fuel assemblies suitable for high neutron flux density environments. Through theoretical derivation of the thermal model of the fuel assembly and sensitivity analysis of the design parameters of the fuel assembly using a single-channel program, the results show that the liquid metal cooling plate fuel element can be well adapted to the high neutron flux Density environment; at relatively low neutron flux densities, bundle fuel elements can also meet reactor design requirements.
{"title":"Optimal Design of Thermal Scheme for Liquid Metal Cooled High Flux Reactor Fuel Assembly","authors":"Rui Pan, Kefan Zhang, Xilin Zhang, Jian Deng, Yong Zhang, D. Zhu, Hongli Chen","doi":"10.1115/icone29-92497","DOIUrl":"https://doi.org/10.1115/icone29-92497","url":null,"abstract":"\u0000 This paper studies the structural parameters of fuel assemblies suitable for high neutron flux density environments.\u0000 Due to the high neutron flux density in the core, high-flux reactors provide an experimental environment for neutron irradiation. However, high neutron flux leads to high heat flux density on the fuel assembly surface, which brings challenges to the design of fuel assembly. Therefore, it is very important to study the structural design of fuel assemblies suitable for high neutron flux density environments.\u0000 Through theoretical derivation of the thermal model of the fuel assembly and sensitivity analysis of the design parameters of the fuel assembly using a single-channel program, the results show that the liquid metal cooling plate fuel element can be well adapted to the high neutron flux Density environment; at relatively low neutron flux densities, bundle fuel elements can also meet reactor design requirements.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131638458","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}
Water is widely used in various nuclear power plants as coolant and moderator due to its good thermal conductivity and stability. However, when the pressure and temperature exceed its critical points, the physical properties of water change drastically, especially the specific heat capacity, which makes the heat transfer phenomenon near the critical point more complicated, even deteriorating. On the other hand, the curved geometry of the spiral pipe leads to a special secondary flow phenomenon in the fluid, which will intensify the heat and mass transfer and improve the heat transfer efficiency. At the same time, the flow will be more complicated in the joint influence of gravity, centrifugal force and Coriolis force. This paper studied the heat transfer characteristics of supercritical water in heated vertical tube by numerical simulation and found that changes of physical properties are the main reason for changes of heat transfer characteristics of supercritical water. And as the heat flux in the tube wall increases, heat transfer deterioration is more prone to occur around the pseudo-critical point. Then it verified the relationship between the criterion of buoyancy and heat transfer deterioration. Last, it explained the mechanism of heat transfer deterioration caused by buoyancy effect by calculating shear stress. In the future, works on the relationship between Bo* and shear stress are expected to carried out.
{"title":"Numerical Analysis on Heat Transfer Characteristics and Buoyancy Effects of Supercritical Pressure Water in Vertical Tube","authors":"Yining Lou, Zhen Zhang","doi":"10.1115/icone29-92951","DOIUrl":"https://doi.org/10.1115/icone29-92951","url":null,"abstract":"\u0000 Water is widely used in various nuclear power plants as coolant and moderator due to its good thermal conductivity and stability. However, when the pressure and temperature exceed its critical points, the physical properties of water change drastically, especially the specific heat capacity, which makes the heat transfer phenomenon near the critical point more complicated, even deteriorating. On the other hand, the curved geometry of the spiral pipe leads to a special secondary flow phenomenon in the fluid, which will intensify the heat and mass transfer and improve the heat transfer efficiency. At the same time, the flow will be more complicated in the joint influence of gravity, centrifugal force and Coriolis force. This paper studied the heat transfer characteristics of supercritical water in heated vertical tube by numerical simulation and found that changes of physical properties are the main reason for changes of heat transfer characteristics of supercritical water. And as the heat flux in the tube wall increases, heat transfer deterioration is more prone to occur around the pseudo-critical point. Then it verified the relationship between the criterion of buoyancy and heat transfer deterioration. Last, it explained the mechanism of heat transfer deterioration caused by buoyancy effect by calculating shear stress. In the future, works on the relationship between Bo* and shear stress are expected to carried out.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"9 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131141142","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}
In this study, we experimentally studied the dynamic response of the sinusoidal single-phase flow in a horizontal narrow rectangular pipe. The experimental results show that, in narrow rectangular channel, the flowrate and the pressure drop also meet the first order dynamic response system phenomenon. The statistic results of prediction by dynamic response system theory shows that, for smaller frequency flow condition, the predictions are precisely, while for higher frequency, the prediction errors are relative bigger, which indicates a research point, high frequency sinusoidal flow, to put effort on.
{"title":"Verification Dynamic Response for Sinusoidal Wave Flow in Narrow Rectangular Channel","authors":"Bao Zhou, Liang Luo, Hongsheng Yuan, P. Gao","doi":"10.1115/icone29-92598","DOIUrl":"https://doi.org/10.1115/icone29-92598","url":null,"abstract":"\u0000 In this study, we experimentally studied the dynamic response of the sinusoidal single-phase flow in a horizontal narrow rectangular pipe. The experimental results show that, in narrow rectangular channel, the flowrate and the pressure drop also meet the first order dynamic response system phenomenon. The statistic results of prediction by dynamic response system theory shows that, for smaller frequency flow condition, the predictions are precisely, while for higher frequency, the prediction errors are relative bigger, which indicates a research point, high frequency sinusoidal flow, to put effort on.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"16 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115336394","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}
Fei Han, Xiting Chen, Yiwu Kuang, Wen Wang, Cheng Ye
Large-scale separate heat pipes used in the complete passive cooling system (PCS) transfer the decay heat in the spent fuel pool (SFP) efficiently through the two-phase natural circulation without any external power. In this study, a lumped mathematical model for the heat pipes are developed and parameters related to the heat transfer ability are discussed to settle the number of the heat pipes under different heat load. For the condensers at the auxiliary building, the effects of the tube pitch and the fin height are discussed, which are key parameters to the heat transfer performance. Different structural designs of the PCS under typical operating conditions are settled. A larger quantity of heat pipes is required for higher decay heat power conditions. To validate the reliability of the PCS, transient three-dimensional simulations of the SFP with immersed evaporators under different heat loads are conducted. Based on the results, detailed thermal-hydraulic characteristics are captured in the pool. Large natural convection circulations are formed at the steady-state. Single flow circulation is formed in the X-Z plane under low heat load cases while a pair of counter-rotate natural convection circulations under high heat load cases. A larger heat load promotes the natural convection intensity and shortens the response time of the PCS. Proper distance between the heat source and heat sink in both vertical and horizontal directions in the SFP is beneficial to the flow organization, improving the heat transfer efficiency of the PCS. The maximum temperature in the SFP is always below the saturation point after the startup of the heat pipes, which could validate the reliability of the PCS and ensure the safety of the plant under emergency conditions.
{"title":"Investigation on a Complete Passive Cooling System Using Large-Scale Separate Heat Pipes in Spent Fuel Pool","authors":"Fei Han, Xiting Chen, Yiwu Kuang, Wen Wang, Cheng Ye","doi":"10.1115/icone29-93413","DOIUrl":"https://doi.org/10.1115/icone29-93413","url":null,"abstract":"\u0000 Large-scale separate heat pipes used in the complete passive cooling system (PCS) transfer the decay heat in the spent fuel pool (SFP) efficiently through the two-phase natural circulation without any external power. In this study, a lumped mathematical model for the heat pipes are developed and parameters related to the heat transfer ability are discussed to settle the number of the heat pipes under different heat load. For the condensers at the auxiliary building, the effects of the tube pitch and the fin height are discussed, which are key parameters to the heat transfer performance. Different structural designs of the PCS under typical operating conditions are settled. A larger quantity of heat pipes is required for higher decay heat power conditions. To validate the reliability of the PCS, transient three-dimensional simulations of the SFP with immersed evaporators under different heat loads are conducted. Based on the results, detailed thermal-hydraulic characteristics are captured in the pool. Large natural convection circulations are formed at the steady-state. Single flow circulation is formed in the X-Z plane under low heat load cases while a pair of counter-rotate natural convection circulations under high heat load cases. A larger heat load promotes the natural convection intensity and shortens the response time of the PCS. Proper distance between the heat source and heat sink in both vertical and horizontal directions in the SFP is beneficial to the flow organization, improving the heat transfer efficiency of the PCS. The maximum temperature in the SFP is always below the saturation point after the startup of the heat pipes, which could validate the reliability of the PCS and ensure the safety of the plant under emergency conditions.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"40 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126903862","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}
If the plant blackout accident occurs in the pool-type sodium-cooled fast reactor, the decay heat of the core is discharged through natural circulation. The Plant Dynamics Test Loop (PLANDTL-DHX), an experimental device built in Japan, can simulate core coolant flow process and decay heat transfer phenomenon under decay heat discharged accident condition. In the present work, the numerical modeling of the experimental device is carried out based on the method of modular modeling and integrated coupling calculation, and the CFD commercial software FLUNET was used for calculation. The rod bundles of different forms in the core are modeled in fine detail. The initial conditions of transient are obtained under the steady boundary condition operation. Then, the change of key thermal parameters such as the temperature of the core and the temperature of IHX inlet and outlet are obtained by simulating the transient accident condition. In addition, there are obvious inner-flow and interflow in the core, meanwhile, the local backflow occurs at the core outlet. The influence of these phenomena on the heat transfer of the whole model is analyzed. The key results of simulation are compared with experimental data. The results can provide numerical references for the discharge of decay heat in the sodium-cooled fast reactor under power blackout accident.
池式钠冷快堆发生核电站停电事故时，堆芯的衰变热通过自然循环排出。核电厂动力学测试回路(Plant Dynamics Test Loop, PLANDTL-DHX)是日本制造的实验装置，可以模拟堆芯冷却剂在衰变放热事故条件下的流动过程和衰变传热现象。在本工作中，基于模块化建模和集成耦合计算的方法对实验装置进行数值建模，并使用CFD商业软件FLUNET进行计算。对岩心中不同形式的棒束进行了详细的建模。在定常边界条件下，得到了暂态的初始条件。然后，通过模拟瞬态事故条件，得到堆芯温度、IHX进出口温度等关键热参数的变化情况。此外，堆芯内部存在明显的内流和互流，同时在堆芯出口处出现局部回流。分析了这些现象对整个模型传热的影响。仿真结果与实验数据进行了比较。研究结果可为停电事故下钠冷快堆衰变热的排放提供数值参考。
{"title":"CFD transient simulation of natural convection characteristics based on detailed core rod bundles model in PLANDTL-DHX experimental device","authors":"Xueyuan Zhang, Yuhao Zhang, Jing Guo, D. Lu","doi":"10.1115/icone29-93169","DOIUrl":"https://doi.org/10.1115/icone29-93169","url":null,"abstract":"\u0000 If the plant blackout accident occurs in the pool-type sodium-cooled fast reactor, the decay heat of the core is discharged through natural circulation. The Plant Dynamics Test Loop (PLANDTL-DHX), an experimental device built in Japan, can simulate core coolant flow process and decay heat transfer phenomenon under decay heat discharged accident condition. In the present work, the numerical modeling of the experimental device is carried out based on the method of modular modeling and integrated coupling calculation, and the CFD commercial software FLUNET was used for calculation. The rod bundles of different forms in the core are modeled in fine detail. The initial conditions of transient are obtained under the steady boundary condition operation. Then, the change of key thermal parameters such as the temperature of the core and the temperature of IHX inlet and outlet are obtained by simulating the transient accident condition. In addition, there are obvious inner-flow and interflow in the core, meanwhile, the local backflow occurs at the core outlet. The influence of these phenomena on the heat transfer of the whole model is analyzed. The key results of simulation are compared with experimental data. The results can provide numerical references for the discharge of decay heat in the sodium-cooled fast reactor under power blackout accident.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"12 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133328260","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 annular fuel element is a safe and efficient new fuel element type. According to previous studies on PWRs, it can bear higher core power density while maintaining or even improving reactor safety when compared to traditional rod lattice fuel rods. In this paper, the application of annular fuel element in lead-cooled fast reactor is studied. Firstly, the flow and heat transfer characteristics of lead-bismuth in annular channel are analyzed by using the computational fluid dynamic method. On this basis, a thermal hydraulic calculation code suitable for annular fuel in lead cooled fast reactor has been developed, which is verified by comparing with CFD results. The application and design optimization of annular fuel in the lead-cooled fast reactor core are carried out using the developed annular fuel single channel code. The effects of different annular fuel parameters on the results are studied, and the core design with rod lattice and annular fuel element is compared and analyzed. The calculation results show that with the same fuel volume and coolant mass flow rate, the maximum core temperature of the annular fuel element is about 600K less than that of the rod bundle fuel element, which shows that the annular fuel element has significant safety superiority.
{"title":"The Thermal-Hydraulic Analysis and Optimization of Annular Fuel for Lead-Cooled Fast Reactor","authors":"Zhang Kefan, Qin Chong, Dong Wenmeng, Pan Rui, Duan Wenshun, Chen Hongli","doi":"10.1115/icone29-92408","DOIUrl":"https://doi.org/10.1115/icone29-92408","url":null,"abstract":"\u0000 The annular fuel element is a safe and efficient new fuel element type. According to previous studies on PWRs, it can bear higher core power density while maintaining or even improving reactor safety when compared to traditional rod lattice fuel rods. In this paper, the application of annular fuel element in lead-cooled fast reactor is studied. Firstly, the flow and heat transfer characteristics of lead-bismuth in annular channel are analyzed by using the computational fluid dynamic method. On this basis, a thermal hydraulic calculation code suitable for annular fuel in lead cooled fast reactor has been developed, which is verified by comparing with CFD results. The application and design optimization of annular fuel in the lead-cooled fast reactor core are carried out using the developed annular fuel single channel code. The effects of different annular fuel parameters on the results are studied, and the core design with rod lattice and annular fuel element is compared and analyzed. The calculation results show that with the same fuel volume and coolant mass flow rate, the maximum core temperature of the annular fuel element is about 600K less than that of the rod bundle fuel element, which shows that the annular fuel element has significant safety superiority.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134115759","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}
Loss of coolant accident (LOCA) is among the important limiting design basis accidents for a PWR nuclear power plant (NPP). In China, a 300MWe two-loop PWR NPP, although facing the challenge of life extension, still adopted rather conservative tools and methods for safety analysis. This is supposed to have guaranteed sufficient margin for safe operation of the plant during the past years, yet, at the expense of plant economy and operation flexibility. To evaluate the safety margin more reasonably and realistically, the mixed methodology of DRHM (deterministic realistic hybrid methodology) is introduced for LBLOCA analysis of the Chinese 300MWe two-loop PWR NPP in the paper, with which conservative evaluation model plus best estimation analysis tool is applied, and effects of uncertainty of important plant state parameters are quantified. In the DRHM analysis of postulated LBLOCA caused by double ended-guillotine-cold-leg break for the 300MWe two-loop PWR NPP in this paper, the evaluation model RELAP5-APK (the conservative Appendix K physical models plus best-estimate system analysis code RELAP5/MOD3) is developed and verified. And during the transient analysis of the LBLOCA scenario, uncertainty of the effects of important plant state parameters are quantified through statistical sampling and corresponding calculation. Taking the cladding peak temperature (PCT) index for demonstration to measure the safety margin, the single-sided confidence upper limit including 95% PCT of the sampling population with 95% confidence level is acquired. The resultant shows that a greater PCT margin is achieved compared with that in the original FSAR. This provide a further confidence for life extension or power uprate of the plant.
{"title":"A Chinese 300MWe Two-Loop PWR NPP LBLOCA Analysis Based on the Deterministic Realistic Hybrid Methodology","authors":"Yuhan Li, B. Kuang","doi":"10.1115/icone29-92431","DOIUrl":"https://doi.org/10.1115/icone29-92431","url":null,"abstract":"\u0000 Loss of coolant accident (LOCA) is among the important limiting design basis accidents for a PWR nuclear power plant (NPP). In China, a 300MWe two-loop PWR NPP, although facing the challenge of life extension, still adopted rather conservative tools and methods for safety analysis. This is supposed to have guaranteed sufficient margin for safe operation of the plant during the past years, yet, at the expense of plant economy and operation flexibility. To evaluate the safety margin more reasonably and realistically, the mixed methodology of DRHM (deterministic realistic hybrid methodology) is introduced for LBLOCA analysis of the Chinese 300MWe two-loop PWR NPP in the paper, with which conservative evaluation model plus best estimation analysis tool is applied, and effects of uncertainty of important plant state parameters are quantified.\u0000 In the DRHM analysis of postulated LBLOCA caused by double ended-guillotine-cold-leg break for the 300MWe two-loop PWR NPP in this paper, the evaluation model RELAP5-APK (the conservative Appendix K physical models plus best-estimate system analysis code RELAP5/MOD3) is developed and verified. And during the transient analysis of the LBLOCA scenario, uncertainty of the effects of important plant state parameters are quantified through statistical sampling and corresponding calculation. Taking the cladding peak temperature (PCT) index for demonstration to measure the safety margin, the single-sided confidence upper limit including 95% PCT of the sampling population with 95% confidence level is acquired. The resultant shows that a greater PCT margin is achieved compared with that in the original FSAR. This provide a further confidence for life extension or power uprate of the plant.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"63 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134119121","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}
Hiroaki Nakanishi, Yoshiteru Komuro, Y. Kondo, Koichi Tanimoto
A two-phase ejector is a device to induce a suction flow without pump or electricity. The flow in the two-phase ejector consists of a drive flow and a suction flow. As the driving flow expands blowing out of a drive flow nozzle, the thermal energy potential is converted into momentum, and by giving it to the suction flow, it is possible to induce the flow without using external power. In a nuclear power plant, a two-phase ejector can be utilized as a device to drive coolant flow in the cases of power failure. Mixing of the drive flow and the suction flow accompanied with evaporation or condensation at the gas-liquid interface depends on thermal hydraulic parameters and flow rate, and it is necessary to control them to maintain the driving force, but it can easily come out of operation range with a slight change in balance. There is little knowledge about heat and mass transfer to find and design operating conditions and ejector configurations. In this study, a heat and mass transfer model of the gas-liquid interface in a critical two-phase flow was developed. To handle thermally non-equilibrium two-phase flow with phase changes occurring simultaneously at the interface, we implemented constitutive equations into CFD tool, such as a correlation for interfacial area concentrations, and we evaluated evaporation coefficient, which is an important parameter to determine the phase change rate, based on the physical property of the working fluid. The CFD simulation method was validated using the experimental data in the literature of a two-phase ejector. In the validation, the flow rates of the drive flow and the suction flow, and pressure distribution inside the ejector were compared. Then, the validity of the developed CFD simulation method have confirmed.
{"title":"Development and Validation of Simulation Method for a Two-Phase Flow Ejector","authors":"Hiroaki Nakanishi, Yoshiteru Komuro, Y. Kondo, Koichi Tanimoto","doi":"10.1115/icone29-92652","DOIUrl":"https://doi.org/10.1115/icone29-92652","url":null,"abstract":"\u0000 A two-phase ejector is a device to induce a suction flow without pump or electricity. The flow in the two-phase ejector consists of a drive flow and a suction flow. As the driving flow expands blowing out of a drive flow nozzle, the thermal energy potential is converted into momentum, and by giving it to the suction flow, it is possible to induce the flow without using external power. In a nuclear power plant, a two-phase ejector can be utilized as a device to drive coolant flow in the cases of power failure.\u0000 Mixing of the drive flow and the suction flow accompanied with evaporation or condensation at the gas-liquid interface depends on thermal hydraulic parameters and flow rate, and it is necessary to control them to maintain the driving force, but it can easily come out of operation range with a slight change in balance. There is little knowledge about heat and mass transfer to find and design operating conditions and ejector configurations.\u0000 In this study, a heat and mass transfer model of the gas-liquid interface in a critical two-phase flow was developed. To handle thermally non-equilibrium two-phase flow with phase changes occurring simultaneously at the interface, we implemented constitutive equations into CFD tool, such as a correlation for interfacial area concentrations, and we evaluated evaporation coefficient, which is an important parameter to determine the phase change rate, based on the physical property of the working fluid. The CFD simulation method was validated using the experimental data in the literature of a two-phase ejector. In the validation, the flow rates of the drive flow and the suction flow, and pressure distribution inside the ejector were compared. Then, the validity of the developed CFD simulation method have confirmed.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"166 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130058396","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 integral test facility is very useful to study the behavior of the pressurized water reactor (PWR) at accidents. As more and more passive safety techniques were adopted in the reactor system, the integral effect test facilities acted very important role to verify these techniques and the prediction of software. An integral effect test facility for the HPR1000 reactor was designed and constructed based on the scaling analysis. The scaling criteria were derived on the model of natural circulation and blowdown of the constant bulk volume in the primary system. The phenomenon were identified and ranked to ensure the scaling can reproduce them in the test facility as the same as the prototype does. The height ratio is 1:4 and the diameter ratio is 1:6 for the test facility. Totally 177 simulators were used to simulate the thermal hydraulics of the fuel assemblies in the practical reactor core. This makes the core keep the same array as the prototype. Each simulator has one electrical heater which power is controlled by the computer. The power of the core has axial cosine profile and three radial zones to reproduce the physical non-uniform distribution in the reactor core.
{"title":"Scaling and Designing Activities of Integral Test Facility for HPR1000 Reactor","authors":"D. Lu, Liangguo Li, Qianhua Su, Jun Xing","doi":"10.1115/icone29-93442","DOIUrl":"https://doi.org/10.1115/icone29-93442","url":null,"abstract":"\u0000 The integral test facility is very useful to study the behavior of the pressurized water reactor (PWR) at accidents. As more and more passive safety techniques were adopted in the reactor system, the integral effect test facilities acted very important role to verify these techniques and the prediction of software. An integral effect test facility for the HPR1000 reactor was designed and constructed based on the scaling analysis. The scaling criteria were derived on the model of natural circulation and blowdown of the constant bulk volume in the primary system. The phenomenon were identified and ranked to ensure the scaling can reproduce them in the test facility as the same as the prototype does. The height ratio is 1:4 and the diameter ratio is 1:6 for the test facility. Totally 177 simulators were used to simulate the thermal hydraulics of the fuel assemblies in the practical reactor core. This makes the core keep the same array as the prototype. Each simulator has one electrical heater which power is controlled by the computer. The power of the core has axial cosine profile and three radial zones to reproduce the physical non-uniform distribution in the reactor core.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125420231","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}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：