International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第9页

Experimental Study of Turbulent Bubbly Flow in 180-Degree Elbow by PIV-PS Technique 180度弯头湍流气泡流动的PIV-PS实验研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.08.003

Wuguang Chen , Songwei Li , Xiaocheng Mu , Fan Yang , Zhigang Yan , Jiajin Sun , Guangyuan Huang , Kangbei Cai , Yuanbiao Hu , Junlian Yin , Dezhong Wang

{"title":"Experimental Study of Turbulent Bubbly Flow in 180-Degree Elbow by PIV-PS Technique","authors":"Wuguang Chen , Songwei Li , Xiaocheng Mu , Fan Yang , Zhigang Yan , Jiajin Sun , Guangyuan Huang , Kangbei Cai , Yuanbiao Hu , Junlian Yin , Dezhong Wang","doi":"10.1016/j.jandt.2022.08.003","DOIUrl":"10.1016/j.jandt.2022.08.003","url":null,"abstract":"<div><p>The curvature effect and bubbles are the key factors that affect the gas-liquid two-phase flow and turbulent characteristics in the 180-degree elbow. Limited by the difficulty of experimental measurement and numerical investigation, little is known about the two-phase flow evolution and turbulent modulation in 180-degree elbow. In this paper, a gas-liquid two-phase measurement technique, simultaneous Particle Image Velocimetry and Pulsed Shadowgraphy (PIV-PS) has been developed by combining two cameras through a spectroscope and optical filters and was applied to explore the two-phase characteristics in 180-degree elbow. In addition, an image post-processing method, based on bubble area detection and remaining particles estimation, was proposed. The characteristics of velocity and turbulent fields in the 180-degree elbow are experimentally studied in different conditions, and the evolution of the flow field in the elbow was also explored in detail. The results show that the horizontal velocity on the centerline of the elbow increases linearly with the liquid flow rates. However, with the introduction of bubbles, the linear feature is eliminated. The turbulent fields show that the introduction of bubbles suppresses turbulence at the low void fraction condition, and enhances at the high void fraction in bubble accumulation areas.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 109-120"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000291/pdfft?md5=33379b29bc382f36fcec55b4c47b4810&pid=1-s2.0-S2468605022000291-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85245917","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Working properties of GEM-TPC at low pressure GEM-TPC的低压工作性能

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.09.002

He Shanhaowei , He Zhixuan , Li Meng , Xiao Chaoyuan , Jiao Weiyao , Yang Herun , Hu Bitao , Zhang Yi

{"title":"Working properties of GEM-TPC at low pressure","authors":"He Shanhaowei , He Zhixuan , Li Meng , Xiao Chaoyuan , Jiao Weiyao , Yang Herun , Hu Bitao , Zhang Yi","doi":"10.1016/j.jandt.2022.09.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2022.09.002","url":null,"abstract":"<div><p>Despite the fact that nuclear technology has already been widely used in the energy sector, the research in this regard have been kept on worldwide. The precise detection of fission fragment tracks is still a knotty issue due to the limits of traditional measurement methods. The inventions of the Gas Electron Multiplier (GEM) as well as its combination with Time Projection Chamber (TPC) detection technology have made many breakthroughs in the multi-parameter correlation measurement of fission fragments tracks. Based on this technology, we designed and built a GEM-TPC detector. To acquire more accurate spatial and energy information, extending the fission fragment tracks by reducing the air pressure in the detector might be an ideal option. At present, since the research on GEM detectors working at low pressure is relatively rare, emphasis in the paper has been put on discussing the working characteristics of our detector in this pressure region in detail. It can be concluded that the lower the working gas pressure, the lower the GEM voltage required to obtain the same gain. In other words, the working voltage range of GEM-TPC will move to a lower level, and the energy resolution of GEM-TPC will first become better and then become worse rapidly as the working gas pressure continues to decrease.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 121-128"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S246860502200031X/pdfft?md5=ea28796d8a4a2a92d9b25b0ed805c40f&pid=1-s2.0-S246860502200031X-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136849091","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Preliminary study on FROBA-SCO2 with transient models and its application for Reactivity Initiated Accident and loss of flow accident 基于暂态模型的FROBA-SCO2及其在反应性引发事故和失流事故中的应用初探

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.08.001

T.A. Yang, K. Zhang, H.K. Yang, W.X. Tian, S.Z. Qiu

{"title":"Preliminary study on FROBA-SCO2 with transient models and its application for Reactivity Initiated Accident and loss of flow accident","authors":"T.A. Yang, K. Zhang, H.K. Yang, W.X. Tian, S.Z. Qiu","doi":"10.1016/j.jandt.2022.08.001","DOIUrl":"10.1016/j.jandt.2022.08.001","url":null,"abstract":"<div><p>As a coolant enabling the reactor with a high efficiency and low volume, the supercritical CO<sub>2</sub> has becoming a hot topic nowadays. However, the fuel rod behaviors under accidents are not studied thoroughly, especially the interaction between the coolant and the cladding. With the previous study of fuel rod behaviors analysis of supercritical CO<sub>2</sub> cooled nuclear reactor under steady state operation, transient state models are implemented and the upgraded code can be used to predict the fuel rod behaviors under RIA and LOFA. After the comparation of the calculation results with SACOS-SCO<sub>2</sub>, the tendency and numerical value are close, so the accuracy is warranted. The upgraded FROBA-SCO<sub>2</sub> is used to predict the fuel rod behaviors under RIA and LOFA conditions and the results indicate that the maximum temperature of cladding and fuel is within safety limits.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 101-108"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000278/pdfft?md5=e807a6461f8f64a85a9d9e2c4d8f0b9e&pid=1-s2.0-S2468605022000278-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73848341","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Simulation study on thermal characteristics of primary coolant system of lead-bismuth cooled reactor 铅铋冷却堆一次冷却剂系统热特性模拟研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.10.002

Shuaijie Lu, Minjun Peng, Genglei Xia, Qianping Zhang

{"title":"Simulation study on thermal characteristics of primary coolant system of lead-bismuth cooled reactor","authors":"Shuaijie Lu, Minjun Peng, Genglei Xia, Qianping Zhang","doi":"10.1016/j.jandt.2022.10.002","DOIUrl":"10.1016/j.jandt.2022.10.002","url":null,"abstract":"<div><p>According to the structural characteristics of the lead-bismuth (LBE) cooled reactor and the flow and heat transfer characteristics of the coolant, a real-time simulation model of the primary coolant system (PCS) is developed. The Reference Fluid Properties (REFPROP) is called to calculate the properties of supercritical carbon dioxide (S–CO<sub>2</sub>), and the International Mathematics Statistics Library (IMSL) is used to call D_IVPAG function to solve the system of stiff differential equations in fully implicit scheme. Using the idea of matrix modularity, the transient code is developed in FORTRAN language. The transient characteristics of power step, primary and secondary flow rate and temperature step are studied respectively. The results show that the transient code can well reflect the dynamic characteristics of the LBE cooled reactor, and the code and algorithm meet the real-time requirements.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 139-146"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000333/pdfft?md5=13898dfb6e993687445e7596885b4db6&pid=1-s2.0-S2468605022000333-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86206319","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development and evaluation of fuel performance analysis code FuSPAC 燃料性能分析代码FuSPAC的开发和评估

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.08.002

Tenglong Cong , Hao Chen , Chao Chen , Shaohong Zhang , Hanyang Gu

{"title":"Development and evaluation of fuel performance analysis code FuSPAC","authors":"Tenglong Cong , Hao Chen , Chao Chen , Shaohong Zhang , Hanyang Gu","doi":"10.1016/j.jandt.2022.08.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2022.08.002","url":null,"abstract":"<div><p>The nuclear fuel rod should maintain self-sustainable nuclear fission, as the first safety barrier to prevent the leakage of radioactive products. The fuel rod performance analyses are essential to ensure reliable long-term operation of fuel rod. In this work, the fuel performance analysis code FuSPAC is developed based on the commonly used 1.5-dimensional method for UO<sub>2</sub> ceramic fuel with Zirconium cladding. The physical and numerical models used in FuSPAC code are described. By comparing the predicted values from FuSPAC code with the available in-reactor fuel rod experimental data and the numerical results from the well-recognized FRAPCON code, the capacity of FuSPAC code on the prediction of fuel pin and cladding under long term irradiation conditions are validated from five aspects, including fission gas release behaviour, thermal behaviour, rod internal void volume, cladding corrosion and cladding hoop strain, showing good accuracy in the behaviour analysis of the irradiated UO<sub>2</sub> fuel rod.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 129-138"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S246860502200028X/pdfft?md5=ed7595e4f394ae35fdb532969adf048f&pid=1-s2.0-S246860502200028X-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"136849090","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Statistical research on 2/3 of the first core nuclear fuel cost in nuclear power plants 核电站1 /3堆芯核燃料成本的统计研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-06-01 DOI: 10.1016/j.jandt.2022.04.001

Zeping Wang , Wenhua Qi , Zongzhi Guo

引用次数: 0

Dimensional analysis on incipient flow instability in one-side high heat loaded rectangular flat heat sink under sub-cooled flow boiling conditions 过冷流动沸腾条件下单侧高热负荷矩形平板散热器初期流动不稳定性的量纲分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-06-01 DOI: 10.1016/j.jandt.2022.03.001

Ji Hwan Lim, Minkyu Park

{"title":"Dimensional analysis on incipient flow instability in one-side high heat loaded rectangular flat heat sink under sub-cooled flow boiling conditions","authors":"Ji Hwan Lim, Minkyu Park","doi":"10.1016/j.jandt.2022.03.001","DOIUrl":"10.1016/j.jandt.2022.03.001","url":null,"abstract":"<div><p>The onset of flow instability (OFI) is a major threat to the operation of a fusion power plant. In particular, because the divertor is loaded with a high heat flux of up to 20 MW/m<sup>2</sup>, it may be vulnerable to OFI. Therefore, in this study, the OFI of a one-side heated flat heat sink was experimentally analyzed. When the effect of the system parameters on the OFI was analyzed, the faster the vapor inside the channel could be removed, the higher was the OFI that was experimentally recorded. As the flow rate and degree of subcooling increase, the OFI increases because it induces an enhancement of the forced convective heat transfer performance and a rapid coding rate, respectively. However, when the pressure is increased, the latent heat and liquid surface tension are reduced; thus, boiling occurs at a lower heat rate, which leads to a decrease in OFI. When the prediction performance of the existing OFI correlations developed under the subcooled flow boiling condition was evaluated, it showed a tendency to overpredict this experimental value significantly. This is because the evaluated correlations were developed based on relatively low heat flux conditions and narrow rectangular channels. Therefore, in this study, we developed a new OFI correlation optimized for experimental values using an artificial intelligence (AI) regression method. Because this correlation can be utilized under one-side high heat load conditions, it is expected that it will be especially useful when establishing the operation strategy of the cooling system of a fusion power plant.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 2","pages":"Pages 43-59"},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000126/pdfft?md5=2309a3682babca963740cfa38b2dc574&pid=1-s2.0-S2468605022000126-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80913035","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Corrigendum to “Pressure drop characteristics in single-side heated circular smooth channel under sub-cooled flow boiling conditions” [Int. J. Adv. Nucl. React. Des. Technol. (2022) 4(2) 60–69] “过冷流动沸腾条件下单面加热圆形光滑通道的压降特性”的勘误表[Int.]。J. Adv.核反应。Des,抛光工艺。(2022) 4(2) 60-69]

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-06-01 DOI: 10.1016/j.jandt.2022.05.003

Ji Hwan Lim, Minkyu Park

引用次数: 0

Pressure drop characteristics in single-side heated circular smooth channel under sub-cooled flow boiling conditions 过冷流动沸腾条件下单侧受热圆形光滑通道压降特性

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-06-01 DOI: 10.1016/j.jandt.2022.03.002

Ji Hwan Lim, Minyku Park

{"title":"Pressure drop characteristics in single-side heated circular smooth channel under sub-cooled flow boiling conditions","authors":"Ji Hwan Lim, Minyku Park","doi":"10.1016/j.jandt.2022.03.002","DOIUrl":"10.1016/j.jandt.2022.03.002","url":null,"abstract":"<div><p>In this study, the two-phase pressure drop for a one-side heated smooth channel was explored through subcooled flow boiling experiments for divertor applications. When the channel heat load was gradually increased, there was little differential pressure change in the single-phase regime, but when it exceeded the onset of significant void (OSV) heat flux, the differential pressure started to increase. It was found that as the subcooling and flow rate increased, the OSV heat flux increased, and the slope of the two-phase differential pressure decreased, This is because it is possible to condense the generated vapor more quickly under these conditions. Moreover, as the pressure increased, the OSV heat flux increased and the size of the ejected single bubble decreased under high pressure conditions. The prediction performances of existing two-phase pressure drop multiplier correlations were also evaluated. Since most of the existing correlations were developed based on channels of a small diameter, and were not evaluated at a high heat flux, that is, in the order of 10 MW/m<sup>2</sup>—they exhibited a high average error rate. Therefore, we developed a new two-phase pressure drop multiplier correlation for a smooth channel under one-side high heat load conditions using PYTHON code and artificial intelligence regression method.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 2","pages":"Pages 60-69"},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000138/pdfft?md5=8cf95173e798e84067008b3b32db6da6&pid=1-s2.0-S2468605022000138-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91449427","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Improvement of a design of the molten salt fast reactor 熔盐快堆设计的改进

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-06-01 DOI: 10.1016/j.jandt.2022.06.001

D.H. Daher , M. Kotb , A.M. Khalaf , Moustafa S. El-Koliel , Abdelfattah Y. Soliman

{"title":"Improvement of a design of the molten salt fast reactor","authors":"D.H. Daher , M. Kotb , A.M. Khalaf , Moustafa S. El-Koliel , Abdelfattah Y. Soliman","doi":"10.1016/j.jandt.2022.06.001","DOIUrl":"10.1016/j.jandt.2022.06.001","url":null,"abstract":"<div><p>This study provides an improved design of molten salt fast reactors to reduce the maximum magnitude velocity in the core. The results indicated that the improved design reduces the magnitude velocity with acceptable hot spot temperature where the maximum magnitude velocity is nearly equal to 4.59 m/s and the temperature of the hot spot is greater than the outlet temperature by 9°. The performed analyses are used to adjust the amount of fissile material started for k<sub>eff</sub> = 1. Neutron flux and delayed Neutron distribution study where the position of the maximum Neutron flux is in the center and the position of the maximum thermal flux is in a fertile blanket. Simulations of unprotected loss of heat sink, unprotected loss of flow, unprotected transient overpower, and Unprotected over-cooling are performed and discussed with new geometry. Due to the negative reactive feedback, it is clear that MSFR has a high safety potential.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 2","pages":"Pages 87-100"},"PeriodicalIF":0.0,"publicationDate":"2022-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000266/pdfft?md5=0b037036065d0d45893a3c56d7491f57&pid=1-s2.0-S2468605022000266-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79711833","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0