Volume 7B: Thermal-Hydraulics and Safety Analysis最新文献_第4页

Critical Heat Flux Experiments for IVR-ERVC Strategy Under the Pool Boiling Condition 池沸腾条件下IVR-ERVC策略的临界热流密度实验

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-93623

G. Wang, B. Kuang, Yihai He, Pengfei Liu

{"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}

引用次数: 0

Optimal Design of Thermal Scheme for Liquid Metal Cooled High Flux Reactor Fuel Assembly 液态金属冷却高通量反应堆燃料组件热方案优化设计

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92497

Rui Pan, Kefan Zhang, Xilin Zhang, Jian Deng, Yong Zhang, D. Zhu, Hongli Chen

引用次数: 0

Numerical Analysis on Heat Transfer Characteristics and Buoyancy Effects of Supercritical Pressure Water in Vertical Tube 超临界压力水垂直管内传热特性及浮力效应数值分析

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92951

Yining Lou, Zhen Zhang

{"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}

引用次数: 0

Verification Dynamic Response for Sinusoidal Wave Flow in Narrow Rectangular Channel 窄矩形通道正弦波流动的动态响应验证

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92598

Bao Zhou, Liang Luo, Hongsheng Yuan, P. Gao

引用次数: 0

Investigation on a Complete Passive Cooling System Using Large-Scale Separate Heat Pipes in Spent Fuel Pool 乏燃料池大型分离热管全被动冷却系统的研究

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-93413

Fei Han, Xiting Chen, Yiwu Kuang, Wen Wang, Cheng Ye

{"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}

引用次数: 0

CFD transient simulation of natural convection characteristics based on detailed core rod bundles model in PLANDTL-DHX experimental device 基于PLANDTL-DHX实验装置岩心棒束详细模型的自然对流特性CFD瞬态模拟

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-93169

Xueyuan Zhang, Yuhao Zhang, Jing Guo, D. Lu

{"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}

引用次数: 0

The Thermal-Hydraulic Analysis and Optimization of Annular Fuel for Lead-Cooled Fast Reactor 铅冷快堆环形燃料的热水力分析与优化

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92408

Zhang Kefan, Qin Chong, Dong Wenmeng, Pan Rui, Duan Wenshun, Chen Hongli

{"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}

引用次数: 0

A Chinese 300MWe Two-Loop PWR NPP LBLOCA Analysis Based on the Deterministic Realistic Hybrid Methodology 基于确定性现实混合方法的中国300MWe双环压水堆NPP LBLOCA分析

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92431

Yuhan Li, B. Kuang

{"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}

引用次数: 0

Development and Validation of Simulation Method for a Two-Phase Flow Ejector 两相流喷射器仿真方法的开发与验证

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-92652

Hiroaki Nakanishi, Yoshiteru Komuro, Y. Kondo, Koichi Tanimoto

{"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}

引用次数: 0

Scaling and Designing Activities of Integral Test Facility for HPR1000 Reactor HPR1000堆整体式试验设备的选型与设计

Volume 7B: Thermal-Hydraulics and Safety Analysis Pub Date : 2022-08-08 DOI: 10.1115/icone29-93442

D. Lu, Liangguo Li, Qianhua Su, Jun Xing

{"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}

引用次数: 0