Subcooled flow boiling in narrow rectangular channel is a very effective heat transfer method and is very important to nuclear reactor design and thermal hydraulic safety. The departure frequency of bubbles in subcooled boiling flow has direct influence on heat transfer. Subcooled flow boiling experiment is carried out to investigate the effect of different working conditions on the bubble departure frequency in the narrow rectangular channel. Because the test section is composed of quartz glass and Sapphire crystal Glass, it can be observed from four different directions to obtain high quality images. The experiment is conducted with mass fluxes of 400 to 1500 kg/(m2s) and heat flow density between 50∼300 (KW/m2). Besides, the system pressure is 100 kPa to 700kpa and inlet subcooled is 10 to 60 °C. Two types of nucleation sites can be detected from the experimental results: sliding growth and condensation. The bubble departure frequency of sliding is significantly higher than that of condensation. Besides, the effect of different working conditions on the bubble departure frequency in narrow rectangular channels was investigated. Finally, the departure frequency is analyzed and compared with previous models. The results show that the model proposed by Cole is in good agreement with the experimental results.
{"title":"Experimental Investigation on the Characteristics of Bubble Departure Frequency of Subcooled Flow Boiling in Narrow Rectangular Channel","authors":"Yong Chen, Hanzhou Liu, D. Chen, Haidong Liu","doi":"10.1115/icone29-93516","DOIUrl":"https://doi.org/10.1115/icone29-93516","url":null,"abstract":"\u0000 Subcooled flow boiling in narrow rectangular channel is a very effective heat transfer method and is very important to nuclear reactor design and thermal hydraulic safety. The departure frequency of bubbles in subcooled boiling flow has direct influence on heat transfer. Subcooled flow boiling experiment is carried out to investigate the effect of different working conditions on the bubble departure frequency in the narrow rectangular channel. Because the test section is composed of quartz glass and Sapphire crystal Glass, it can be observed from four different directions to obtain high quality images. The experiment is conducted with mass fluxes of 400 to 1500 kg/(m2s) and heat flow density between 50∼300 (KW/m2). Besides, the system pressure is 100 kPa to 700kpa and inlet subcooled is 10 to 60 °C. Two types of nucleation sites can be detected from the experimental results: sliding growth and condensation. The bubble departure frequency of sliding is significantly higher than that of condensation. Besides, the effect of different working conditions on the bubble departure frequency in narrow rectangular channels was investigated. Finally, the departure frequency is analyzed and compared with previous models. The results show that the model proposed by Cole is in good agreement with the experimental results.","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":"131219866","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}
Ji Wenying, Lu Changdong, Ou Yangyong, Chen Yichen
To evaluate the thermal performance of a closed-loop Passive Containment Cooling System (PCCS) based on hot pipe technology under nuclear accidents, the coupled effect of three systems of reactor primary cooling circuit, containment and the PCCS is analyzed under LB-LOCA with SBO and SB-LOCA with SBO with the PCCS started one hour after the accident. The coupled analysis method is that the heat transfer power of PCCS simulated by the LOCUST model and the mass and energy releasing results obtained from the integrated analysis program for serious accidents are the inputs for containment thermal response analysis code. The results showed that in the 0∼10 hours after the accident, the containment pressure is less than the design limit of 5.20bar.a; in the middle and long term of ten more hours after the accident, the containment pressure is less than 3.50bar.a, the integrity of the containment can be ensured and the feasibility of the design of the PCCS in this paper is validated.
{"title":"Analysis of the Thermal Performance of the Passive Containment Cooling System Based on Heat Pipe Technology Under Nuclear Accidents","authors":"Ji Wenying, Lu Changdong, Ou Yangyong, Chen Yichen","doi":"10.1115/icone29-93273","DOIUrl":"https://doi.org/10.1115/icone29-93273","url":null,"abstract":"\u0000 To evaluate the thermal performance of a closed-loop Passive Containment Cooling System (PCCS) based on hot pipe technology under nuclear accidents, the coupled effect of three systems of reactor primary cooling circuit, containment and the PCCS is analyzed under LB-LOCA with SBO and SB-LOCA with SBO with the PCCS started one hour after the accident. The coupled analysis method is that the heat transfer power of PCCS simulated by the LOCUST model and the mass and energy releasing results obtained from the integrated analysis program for serious accidents are the inputs for containment thermal response analysis code. The results showed that in the 0∼10 hours after the accident, the containment pressure is less than the design limit of 5.20bar.a; in the middle and long term of ten more hours after the accident, the containment pressure is less than 3.50bar.a, the integrity of the containment can be ensured and the feasibility of the design of the PCCS in this paper is validated.","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":"133237959","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}
Liquid lead or lead-bismuth eutectic has good thermal physical properties, neutron performances, and natural circulation capacity. It is the first choice of coolant material for the lead-based fast reactor (LFR). However, there is a fundamental problem, heat-transfer characteristics and quantitative relationship of liquid lead and lead-bismuth eutectic, to be solved in the process of thermal-hydraulic design and safety analysis code development of lead-based fast reactor. The existing correlations are mostly fitted by the experimental data under specific conditions and the applicability is poor in a wide range of parameters, which affects the accuracy of thermal-hydraulic design and safety analysis results. Therefore, it is a necessity to establish experimental databases and correlation databases to obtain a more accurate heat transfer correlation of liquid lead and bismuth eutectics within a wider range of parameters. This paper reviews the typical theoretical and experimental studies on the heat transfer characteristics of liquid lead and lead-bismuth eutectics in the recent 70 years, and an experimental database based on the heat transfer experimental data in the published literature is established, which data points include a wide range of thermodynamic parameters — inner diameter ranging from 9.00 to 16.56 mm, temperature from 250.0 to 412.80 °C, heat fluxes from 20.0 to 12900.0 kW/m2, and mass fluxes from 0 to 12.0 kg/m2s. Based on the database, the typical heat transfer correlations of liquid lead and lead-bismuth eutectics are evaluated and analyzed, and a new heat transfer correlation is developed, its suitable range is 100 < Pe < 5800. The results show that the mean absolute relative error of the new correlation is 16.64%, the average relative error is 0.74%, the standard deviation is 4.37%, and the error of 65.27% of the total data points is within ± 20%. Therefore, the heat-transfer correlation proposed in this study can well predict the heat-transfer coefficient of high-temperature liquid lead or lead-bismuth eutectic in the tube.
液态铅或铅铋共晶具有良好的热物理性能、中子性能和自然循环能力。它是铅基快堆(LFR)的首选冷却材料。然而，在铅基快堆的热工设计和安全分析规范制定过程中，存在着一个基本问题，即液态铅和铅铋共晶的传热特性及其定量关系。现有的关系式大多是用特定条件下的实验数据拟合的，在较宽的参数范围内适用性较差，影响了热工设计和安全分析结果的准确性。因此，有必要建立实验数据库和相关数据库，以便在更大的参数范围内获得更准确的液态铅铋共晶传热相关性。本文回顾了近70年来关于液态铅和铅铋共晶传热特性的典型理论和实验研究，并基于已发表文献的传热实验数据建立了一个实验数据库，该数据库的数据点包括广泛的热力学参数内径范围为9.00至16.56 mm，温度范围为250.0至412.80°C，热流从20.0至12900.0 kW/m2，质量通量从0至12.0 kg/m2。在此基础上，对液态铅和铅铋共晶的典型传热关系式进行了评价和分析，建立了一个新的传热关系式，其适用范围为100 < Pe < 5800。结果表明，新相关性的平均绝对相对误差为16.64%，平均相对误差为0.74%，标准差为4.37%，总数据点误差在±20%以内的误差为65.27%。因此，本文提出的传热相关性可以很好地预测管内高温液态铅或铅铋共晶的传热系数。
{"title":"A New Correlation for Predictions of Heat Transfer of Lead-Bismuth Eutectic in Circular Tubes","authors":"Zhengquan Wang, Q. Wen, Shijia Xu","doi":"10.1115/icone29-92378","DOIUrl":"https://doi.org/10.1115/icone29-92378","url":null,"abstract":"\u0000 Liquid lead or lead-bismuth eutectic has good thermal physical properties, neutron performances, and natural circulation capacity. It is the first choice of coolant material for the lead-based fast reactor (LFR). However, there is a fundamental problem, heat-transfer characteristics and quantitative relationship of liquid lead and lead-bismuth eutectic, to be solved in the process of thermal-hydraulic design and safety analysis code development of lead-based fast reactor. The existing correlations are mostly fitted by the experimental data under specific conditions and the applicability is poor in a wide range of parameters, which affects the accuracy of thermal-hydraulic design and safety analysis results. Therefore, it is a necessity to establish experimental databases and correlation databases to obtain a more accurate heat transfer correlation of liquid lead and bismuth eutectics within a wider range of parameters.\u0000 This paper reviews the typical theoretical and experimental studies on the heat transfer characteristics of liquid lead and lead-bismuth eutectics in the recent 70 years, and an experimental database based on the heat transfer experimental data in the published literature is established, which data points include a wide range of thermodynamic parameters — inner diameter ranging from 9.00 to 16.56 mm, temperature from 250.0 to 412.80 °C, heat fluxes from 20.0 to 12900.0 kW/m2, and mass fluxes from 0 to 12.0 kg/m2s. Based on the database, the typical heat transfer correlations of liquid lead and lead-bismuth eutectics are evaluated and analyzed, and a new heat transfer correlation is developed, its suitable range is 100 < Pe < 5800. The results show that the mean absolute relative error of the new correlation is 16.64%, the average relative error is 0.74%, the standard deviation is 4.37%, and the error of 65.27% of the total data points is within ± 20%. Therefore, the heat-transfer correlation proposed in this study can well predict the heat-transfer coefficient of high-temperature liquid lead or lead-bismuth eutectic in the tube.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"14 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":"128788502","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}
Passive residual heat removal heat exchanger (PRHR HX) is one of the important equipment in passive residual heat removal system (PRHRs) of marine nuclear power platform. The research on the condensation heat transfer characteristics of saturated steam in PRHR HX tube under tilting condition can provide support and optimization for passive safety system design of marine nuclear power platform. The work in this paper is of great significance to safe operation of marine nuclear power platform. The heat transfer characteristics of saturated steam condensation in the PRHR HX tube under tilting condition are analyzed by building an experimental device with a power ratio of 1:50. The experimental results show that within the range of experimental parameters in this paper, compared with the static state of the experimental device, the condensation heat transfer coefficient of saturated steam in PRHR HX tube under tilting condition is increased. When the heat flux is 190kW/m2, the heat transfer coefficient of saturated steam condensation in PRHR HX tube under tilting condition increases by about 40% compared with static condition. In this paper the formula for calculating the condensation heat transfer coefficient of saturated steam in PRHR HX tube is revised by introducing the tilt angle. The relative error between the modified formula and the experimental value is within ±10%. The research results of this paper can provide reference for the design and optimization of passive safety systems for marine nuclear power platform and similar applications.
{"title":"Research on Characteristics of Condensation Heat Transfer for Marine Nuclear Power Platform PRHR HX Under Tilting Condition","authors":"Pengzheng Li, Yongquan Li, Shaoyou Liu, Dong Zhu, Zhiqiang Zhu, Xiaming Kong","doi":"10.1115/icone29-92406","DOIUrl":"https://doi.org/10.1115/icone29-92406","url":null,"abstract":"\u0000 Passive residual heat removal heat exchanger (PRHR HX) is one of the important equipment in passive residual heat removal system (PRHRs) of marine nuclear power platform. The research on the condensation heat transfer characteristics of saturated steam in PRHR HX tube under tilting condition can provide support and optimization for passive safety system design of marine nuclear power platform. The work in this paper is of great significance to safe operation of marine nuclear power platform. The heat transfer characteristics of saturated steam condensation in the PRHR HX tube under tilting condition are analyzed by building an experimental device with a power ratio of 1:50. The experimental results show that within the range of experimental parameters in this paper, compared with the static state of the experimental device, the condensation heat transfer coefficient of saturated steam in PRHR HX tube under tilting condition is increased. When the heat flux is 190kW/m2, the heat transfer coefficient of saturated steam condensation in PRHR HX tube under tilting condition increases by about 40% compared with static condition. In this paper the formula for calculating the condensation heat transfer coefficient of saturated steam in PRHR HX tube is revised by introducing the tilt angle. The relative error between the modified formula and the experimental value is within ±10%. The research results of this paper can provide reference for the design and optimization of passive safety systems for marine nuclear power platform and similar applications.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"33 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":"131687114","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 10kWe integrated reactor with Stirling generator is in design, to satisfy China’s power demand for both Earth orbit and deep space exploration in the next two decades. The integration of the core and the thermoelectric conversion system, reduces the number of tubes and pump structures, which leads to a higher energy conversion rate, fewer failure risks and coolant leaks. The waste heat of this reactor would be transferred through its heat pipes to its radiators, then to the space. Applying nanofluids would help reduce the heat pipe sizes, because nanofluids have great alterations in their thermo-physical properties with a small fraction of nanoparticles. Numerous models have been proposed to characterize the thermo-physical properties of nanofluids. However, it is found that researchers have different, sometimes even contradictory conclusions about some of the properties. At the same time, these properties could be affected by various aspects, the simple models are not sufficient for the reference. This work focuses on evaluating the models of density, specific heat capacity, thermal conductivity, viscosity, and Nusselt number of nanofluids with statistical methods, and provides reference thermo-physical properties for the design of the heat pipes in the space reactor. For this reason, a great amount of experimental data is collected. Profiting from the collected data, artificial neural network (ANN) models based on Pytorch are trained and compared with the other models.
{"title":"Model Evaluation of Various Thermo-Physical Properties of Nanofluids and ANN Modelling for 10kWe Integrated Reactor","authors":"Lingyun Zheng, Zhi-gang Zhang, Xin-wen Wang","doi":"10.1115/icone29-92476","DOIUrl":"https://doi.org/10.1115/icone29-92476","url":null,"abstract":"\u0000 A 10kWe integrated reactor with Stirling generator is in design, to satisfy China’s power demand for both Earth orbit and deep space exploration in the next two decades. The integration of the core and the thermoelectric conversion system, reduces the number of tubes and pump structures, which leads to a higher energy conversion rate, fewer failure risks and coolant leaks. The waste heat of this reactor would be transferred through its heat pipes to its radiators, then to the space. Applying nanofluids would help reduce the heat pipe sizes, because nanofluids have great alterations in their thermo-physical properties with a small fraction of nanoparticles. Numerous models have been proposed to characterize the thermo-physical properties of nanofluids. However, it is found that researchers have different, sometimes even contradictory conclusions about some of the properties. At the same time, these properties could be affected by various aspects, the simple models are not sufficient for the reference. This work focuses on evaluating the models of density, specific heat capacity, thermal conductivity, viscosity, and Nusselt number of nanofluids with statistical methods, and provides reference thermo-physical properties for the design of the heat pipes in the space reactor. For this reason, a great amount of experimental data is collected. Profiting from the collected data, artificial neural network (ANN) models based on Pytorch are trained and compared with the other models.","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":"128903050","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}
Xianbing Chen, Chengyi Long, Dongbao Zhu, P. Gao, Xiaming Kong, Lu Yao
Nuclear power in a water-cooled reactor changes with reactivity under the influence of temperature and void fraction. Effects of power oscillations on natural circulation are experimentally investigated with a neutronic-thermo-hydraulic loop to better understand neutronic-thermo-hydraulic effects. Heating power can be precisely controlled by a DC power supply. Both stable natural circulation flow and flow instability experiments are conducted. Amplitude of flow oscillations increases with the increase of amplitude of power oscillations for the stable natural circulation flow. Period and amplitude of power oscillations are compared with those of flow oscillations. Period of flow oscillations corresponds with the period of power oscillations due to the relative balance between driving force and resistance. Amplitude and period of power oscillations are changed to obtain stability boundary by keeping other parameters constant. Experimental results indicated that small amplitude and short period power oscillations has little influence of the stability boundary. However, power oscillations cause the premature of flow instability when period and amplitude of power oscillations are further increased. The destabilizing effects of power oscillations on natural circulation flow depend on the period and amplitude. The stability region reduces with the increase of amplitude and period of power oscillation in this experimental loop.
{"title":"Effects of Power Oscillations on Natural Circulation Flow Instability With a Neutronic-Thermo-Hydraulic Loop","authors":"Xianbing Chen, Chengyi Long, Dongbao Zhu, P. Gao, Xiaming Kong, Lu Yao","doi":"10.1115/icone29-93303","DOIUrl":"https://doi.org/10.1115/icone29-93303","url":null,"abstract":"\u0000 Nuclear power in a water-cooled reactor changes with reactivity under the influence of temperature and void fraction. Effects of power oscillations on natural circulation are experimentally investigated with a neutronic-thermo-hydraulic loop to better understand neutronic-thermo-hydraulic effects. Heating power can be precisely controlled by a DC power supply. Both stable natural circulation flow and flow instability experiments are conducted. Amplitude of flow oscillations increases with the increase of amplitude of power oscillations for the stable natural circulation flow. Period and amplitude of power oscillations are compared with those of flow oscillations. Period of flow oscillations corresponds with the period of power oscillations due to the relative balance between driving force and resistance. Amplitude and period of power oscillations are changed to obtain stability boundary by keeping other parameters constant. Experimental results indicated that small amplitude and short period power oscillations has little influence of the stability boundary. However, power oscillations cause the premature of flow instability when period and amplitude of power oscillations are further increased. The destabilizing effects of power oscillations on natural circulation flow depend on the period and amplitude. The stability region reduces with the increase of amplitude and period of power oscillation in this experimental loop.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"59 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":"128630441","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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