International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第8页

Influence analysis of alloy elements on irradiation embrittlement of RPV steel based on deep neural network 基于深度神经网络的合金元素对RPV钢辐照脆化的影响分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2023-03-01 DOI: 10.1016/j.jandt.2023.03.002

Bai Bing, Xu Han, Lixia Jia, Xinfu He, Changyi Zhang, Wen Yang

{"title":"Influence analysis of alloy elements on irradiation embrittlement of RPV steel based on deep neural network","authors":"Bai Bing, Xu Han, Lixia Jia, Xinfu He, Changyi Zhang, Wen Yang","doi":"10.1016/j.jandt.2023.03.002","DOIUrl":"https://doi.org/10.1016/j.jandt.2023.03.002","url":null,"abstract":"<div><p>Reactor pressure vessel (RPV) is the most important core equipment in PWR. Its service life determines the service life of nuclear power plant and directly affects the economy and safety of nuclear power plant. Because RPV is serviced at high temperature, high pressure and high energy neutrons for a long time, the properties of RPV steel will significantly degrade, in which irradiation embrittlement is the most important factor for the structural integrity of RPV. In this work, about 700 groups of data such as composition, irradiation conditions and ductile brittle transition temperature of RPV steel are collected, and the data are cleaned and screened for modelling by machine learning. The deep neural network is used for establishing the correlation between key component and irradiation embrittlement of RPV steel. The results show that the lower flux of neutron irradiation will make the radiation embrittlement effect of RPV steel more obvious at the same neutron fluence. Cu, P and Ni are the key factors to influence the △DBTT of RPV steel. The synergistic effect of Cu and Ni on irradiation embrittlement is greater than that of Cu and Mn. These results will help to promote the optimization design of new RPV steel.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 1","pages":"Pages 44-51"},"PeriodicalIF":0.0,"publicationDate":"2023-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49748065","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: 1

Characteristic analysis of natural circulation residual heat removal in small reactor 小型反应器自然循环除余热特性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.12.001

Haoyang Liao , Xianbo Wang , Jia Ma , Fulong Zhao , Chengtao Li , Sichao Tan

{"title":"Characteristic analysis of natural circulation residual heat removal in small reactor","authors":"Haoyang Liao , Xianbo Wang , Jia Ma , Fulong Zhao , Chengtao Li , Sichao Tan","doi":"10.1016/j.jandt.2022.12.001","DOIUrl":"10.1016/j.jandt.2022.12.001","url":null,"abstract":"<div><p>In order to study the characteristics of small helium-xenon cooled nuclear reactor that only relies on natural circulation to remove residual heat after shutdown, this paper carries out geometric and physical modeling of helium-xenon cooled nuclear reactor, and uses CFD to simulate the process of natural circulation residual heat removal. Variation laws of the hot spot temperature on the surface of fuel cladding and the mass flow of helium-xenon mixed gas under different working conditions are calculated, and the transient characteristics of temperature, flow and other related parameters in the system loop are analyzed. The calculation results show that during the natural circulation residual heat removal process, the overall change trend of the hot spot temperature on the surface of the fuel cladding, the average temperature of the primary coolant and reactor outlet temperature first rises to the highest value and then decreases, and the mass flow of helium-xenon mixed gas shows an exponential decay trend; Increasing the initial mass flow of helium-xenon mixed gas and reducing the residual heat of the reactor core can reduce the maximum hot spot temperature on the surface of the fuel cladding. The relevant research results provide a useful reference for the optimal design of natural circulation residual heat removal of helium-xenon cooled nuclear reactor.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 187-195"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S246860502200059X/pdfft?md5=0432f8c152d440b116025830d6efe4b7&pid=1-s2.0-S246860502200059X-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80453720","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

Experimental investigation of nano-particle deposited wick Structure's heat transfer characteristics 纳米颗粒沉积芯结构传热特性的实验研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.12.003

Ruiyu Han, Zetao Wang, Kailun Guo, Chenglong Wang, Dalin Zhang, Wenxi Tian, Suizheng Qiu, Guanghui Su

{"title":"Experimental investigation of nano-particle deposited wick Structure's heat transfer characteristics","authors":"Ruiyu Han, Zetao Wang, Kailun Guo, Chenglong Wang, Dalin Zhang, Wenxi Tian, Suizheng Qiu, Guanghui Su","doi":"10.1016/j.jandt.2022.12.003","DOIUrl":"10.1016/j.jandt.2022.12.003","url":null,"abstract":"<div><p>Recently, nanotechnology attracts more and more attention in heat transfer system. Many studies have shown that using nanofluid as the working fluid can improve the thermal performance of the heat pipe. In this paper, the preliminary effort of applying nano-surface engineering on the heat pipe wick to fabricate the nano-structured wick structure has been carried out. Experimental investigation has been performed to determine the effect of the heat source power, mesh aperture and the wick surface structure on the total thermal resistance of the wick structure. When the heat source power increased from 20 W to 30 W, then to 50 W, the evaporation resistance decreases 25.9% and 13.7% respectively, while the conduction resistance increases 24.8% at first and then decreases 9.5%. The evaporation resistance is less with than without the wick structure, and it increases with the increase of mesh aperture. When the mesh aperture increases from 75 μm to 150 μm, then to 460 μm, the evaporation resistance increases 1.9% and 1.7% respectively. Whereas, the wick structure, as well as the mesh aperture have slight influence on the conduction resistance. The experimental results suggest that the nano structure on the surface can efficiently decrease the evaporation resistance but increase the conduction resistance at the same time. Compared to the wick structure with smooth surface, the evaporation resistance of nano-surface deposited by 0.01% TiO<sub>2</sub> and 0.02% TiO<sub>2</sub> has decreased 4.3% and 15.5%, respectively. This study provides a reference for the preparation of the heat pipe with greater performance in the future.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 196-204"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000618/pdfft?md5=dad5647d895d3fd25dcc8e429ebeec25&pid=1-s2.0-S2468605022000618-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91505127","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: 2

Heat transfer performance of high temperature and high velocity hydrogen flow and analysis of blockage characteristics 高温高速氢气流的传热性能及堵塞特性分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.12.002

Wenxuan Ju, Xianmin Dong, Ruibo Lu, Haoran Shen, Fulong Zhao, Sichao Tan

{"title":"Heat transfer performance of high temperature and high velocity hydrogen flow and analysis of blockage characteristics","authors":"Wenxuan Ju, Xianmin Dong, Ruibo Lu, Haoran Shen, Fulong Zhao, Sichao Tan","doi":"10.1016/j.jandt.2022.12.002","DOIUrl":"10.1016/j.jandt.2022.12.002","url":null,"abstract":"<div><p>In the current situation of increasing demand for nuclear heat propulsion in various countries, the heat transfer characteristics of high temperature and high velocity fluid in the flow channel are particularly important for safe core operation and efficient heat transfer. In order to obtain the flow and heat transfer characteristics of the coolant channel of nuclear thermal propulsion reactor under normal and blocked flow conditions, the numerical method based on commercial CFD software is adopted. Researches by changing the inlet velocity of the coolant channel and the blocking forms of the circular pipe are carried out. The SST k-omega viscous model together with the pressure-based coupling algorithm, which have extraordinary applicability to the model, are adopted. The impulse which are vital to the performance of nuclear heat propulsion is analyzed.</p><p>The calculation results indicate that with the increase of the inlet velocity, the coolant outlet impulse increases significantly; The heat transfer deteriorated gradually, and the wall temperature slightly increased; The closer to the center of the pipe the blockage occurred, the higher the pipe temperature became. The research methods and results can provide a reference for the study of the heat transfer characteristics of high temperature and high velocity hydrogen flow, and the heat transfer characteristics and transient simulation initiated by the blockage of the nuclear thermal propulsion reactor.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 205-216"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000606/pdfft?md5=46be7b7e96de7f98df2d8b4fa441b42a&pid=1-s2.0-S2468605022000606-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78893256","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 of an edge-cloud collaboration framework for fission battery management system 裂变电池管理系统边缘云协作框架的开发

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.11.002

Hong Xu , Yihua Duo , Tao Tang

{"title":"Development of an edge-cloud collaboration framework for fission battery management system","authors":"Hong Xu , Yihua Duo , Tao Tang","doi":"10.1016/j.jandt.2022.11.002","DOIUrl":"10.1016/j.jandt.2022.11.002","url":null,"abstract":"<div><p>As a “plug-and-play” nuclear system, the technology of fission battery enables nuclear reactor systems to function as batteries. Undoubtedly, advanced intelligent technologies have to be adopted to realize the excellent attributes of fission batteries, i.e., economic, standardized, installed, unattended, and reliable, and its commercial applications. This article proposed the edge-cloud collaboration technology for fission batteries to satisfy its intelligent requirements. A detailed framework for a fission battery management system has been developed. To develop such a management system, four types of concrete transmission information for the edge-cloud collaboration (i.e. resource and data, AI technologies, application and services, safety and strategies) are identified. The edge and cloud designs are also discussed in detail, including their hardware and software. A functional framework of edge-cloud collaboration network platform for fission battery management system is developed preliminarily. The framework is based on a three-layer structure, i.e., perception layer, technical layer and application layer. Although no detailed algorithm or verification is introduced (since the edge-cloud collaboration framework has been widely used in some industry fields but not in nuclear engineering), as primarily work for the future detailed study of fission battery, it is necessary and extremely important.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 177-186"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000588/pdfft?md5=031534d7d51a44f7fee0621d3a7211b2&pid=1-s2.0-S2468605022000588-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79739895","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

Melting and relocation behavior of metal and force balance analysis of metallic droplet 金属的熔化和重定位行为及金属液滴的力平衡分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.12.004

Kui Ge, Yapei Zhang, Kui Zhang, Wenxi Tian, G.H. Su, Suizheng Qiu

{"title":"Melting and relocation behavior of metal and force balance analysis of metallic droplet","authors":"Kui Ge, Yapei Zhang, Kui Zhang, Wenxi Tian, G.H. Su, Suizheng Qiu","doi":"10.1016/j.jandt.2022.12.004","DOIUrl":"10.1016/j.jandt.2022.12.004","url":null,"abstract":"<div><p>In this paper, the heating, melting and relocation behavior of zinc was carried out based on the MEREA (Metal mElting and RElocation progression Apparatus) facility. Metal melt tests with different axial power distributions were conducted. The visualizing data, temperature profiles and parameters of the relocation process were obtained. As the temperature rises, the metal melted and relocated downward. The melt solidified in lower colder parts and formed melt accumulations. The increasing volume of accumulations causes some bulges on the metal surface and further cracks. Then melt flowed out through the cracks and aggregated into some metallic droplets. As the droplet volume increases, it relocated downward as a rivulet driven by gravity. Besides, some droplets might depart at the surface and fall into the lower head in an approximately free fall. The onset of melt droplet motion (slip or departure) was analyzed based on the force balance method. Some important model parameters and criteria of the droplet motion were also obtained based on the MEREA experimental data. The results of this research can provide a better understanding of the relocation behavior of metal and offer references for model development and verification.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 217-226"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605023000017/pdfft?md5=cc4962f5d6b4c38cccca621408832e44&pid=1-s2.0-S2468605023000017-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90064937","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

Electrochemical behavior and corrosion rate prediction study of alloy 690 690合金的电化学行为及腐蚀速率预测研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-12-01 DOI: 10.1016/j.jandt.2022.11.001

Weibing Wang , Ying Zhou , Bo Liang , Bo Wang , Meng Zhang , Sichao Tan

{"title":"Electrochemical behavior and corrosion rate prediction study of alloy 690","authors":"Weibing Wang , Ying Zhou , Bo Liang , Bo Wang , Meng Zhang , Sichao Tan","doi":"10.1016/j.jandt.2022.11.001","DOIUrl":"10.1016/j.jandt.2022.11.001","url":null,"abstract":"<div><p>Alloy 690 is a material commonly used in heat transfer tubes of steam generators (SGTs). The electrochemical corrosion behavior in different volumes (<em>V</em>) of NaCl–Na<sub>2</sub>S<sub>2</sub>O<sub>3</sub> solution for different surface roughness (<em>R</em><sub>a</sub>) of alloy 690 was studied, and a BPNN model for predicting electrochemical corrosion parameters of alloy 690 was established (CP and CCD). The results show that <em>R</em><sub>a</sub> on the surface of alloy 690 destroys the passivation film and changes the oxygen vacancies in the passivation film, so that Cl<sup>−</sup> is more easily attached to the passivation film. Finally, accelerated electrochemical reactions on the surface of SGTs. When the solution is Na<sub>2</sub>S<sub>2</sub>O<sub>3</sub>, with the increase of <em>R</em><sub>a</sub> on the metal surface (0.0977≤ <em>R</em><sub>a</sub> ≤ 1.1661), the corrosion potential gradually decreases. BPNN can accurately predict the electrochemical corrosion rate of alloy 690 under different conditions, and the relative error is less than 10%. In this paper, an artificial intelligence model is established for the first time to predict the relationship between <em>R</em><sub>a</sub> and electrochemical corrosion, and the conclusion can be used to schedule the maintenance process of the SG, thereby reducing the risk of structural failure and maintenance costs.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 4","pages":"Pages 171-176"},"PeriodicalIF":0.0,"publicationDate":"2022-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000576/pdfft?md5=3b7da9cc84fd91d10b5cf38192844dd4&pid=1-s2.0-S2468605022000576-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88521409","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: 6

Reliability assessment methods to address fast transient of reactor core 处理堆芯快速瞬变的可靠性评估方法

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.09.001

N.H. Badrun , Nandita Talukder , Nosrat Sharmin

{"title":"Reliability assessment methods to address fast transient of reactor core","authors":"N.H. Badrun , Nandita Talukder , Nosrat Sharmin","doi":"10.1016/j.jandt.2022.09.001","DOIUrl":"10.1016/j.jandt.2022.09.001","url":null,"abstract":"<div><p>In order to enhance the safety of new advanced reactors, reliability based approach to the design of thermal hydraulic system becomes necessary. In this work, “load exceeds capacity” based approach of structural reliability analysis is employed and probability of failure of the system was then assessed in terms of a limit state function while probabilistic measure of limit state function violation is performed through different methods of reliability assessment. Here, we have focused on TRIGA core subjected to reactivity initiated fast transient. Initially, response surface design method has been used for approximating true failure surface, and then FORM-SORM analysis has been carried out. But, due to non linearity involved with failure surface, there have been noticed instability in FORM-SORM implementation. Later, directional simulation approach of Monte Carlo variance reduction techniques has been employed to illustrate such fast transient. In the investigation, there have been several aspects considered and in each case directional simulation method has shown its ability to give valid results. Hence, the method could be recommended as a viable and efficient scheme to solve even fast transient problem in design and analysis of any reactor.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 156-163"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000308/pdfft?md5=0f6e85b4cda0f184fc2ed07d3b18c549&pid=1-s2.0-S2468605022000308-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88343092","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

Numerical study of gas-injection induced pool sloshing behavior using MPS method 注气诱导池晃动行为的MPS数值研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.10.001

Junru Lv , Xiaoxing Liu , Kai Wang , Songbai Cheng , Lili Tong

{"title":"Numerical study of gas-injection induced pool sloshing behavior using MPS method","authors":"Junru Lv , Xiaoxing Liu , Kai Wang , Songbai Cheng , Lili Tong","doi":"10.1016/j.jandt.2022.10.001","DOIUrl":"10.1016/j.jandt.2022.10.001","url":null,"abstract":"<div><p>When a Core Disruptive Accident (CDA) occurs in a Sodium-cooled Fast Reactor (SFR), as the accident progresses, its core may form a pool of molten fuel. When the molten fuel pool expands, a certain amount of coolant may be entrained in it. Because of the Fuel-Coolant Interaction (FCI), the molten fuel pool will have centralized sloshing behavior, making the fuel distribution more dense, leading to the risk of re-criticality, therefore, the study of centralized sloshing behavior is of great significance for evaluating the impact of the CDA. Under various experimental conditions, the sloshing experiment of gas injection can effectively simulate the mechanism and characteristics of centralized sloshing behavior. In this study, numerical simulation of gas-injection induced sloshing behavior is performed based on the Moving Particle Semi-implicit (MPS) method. Simulation results are compared with the experimental results. The results show that the deformation of bubble shape and liquid level in the sloshing characteristics are in good agreement with the experimental results. Therefore, the numerical simulation method established in this study can be used to study the pool sloshing characteristics induced by gas injection.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 147-155"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000321/pdfft?md5=446d8bff407779dbfc824cb7e254c8ca&pid=1-s2.0-S2468605022000321-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83631995","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: 2

Investigation on radial fuel relocation and its influence on heat split phenomenon of dual-cooled annular fuel element 双冷环形燃料元件径向重新安置及其对热分裂现象影响的研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI: 10.1016/j.jandt.2022.10.003

Yangbin Deng, Zijian Xu, Yuan Yin

{"title":"Investigation on radial fuel relocation and its influence on heat split phenomenon of dual-cooled annular fuel element","authors":"Yangbin Deng, Zijian Xu, Yuan Yin","doi":"10.1016/j.jandt.2022.10.003","DOIUrl":"10.1016/j.jandt.2022.10.003","url":null,"abstract":"<div><p>Radial fuel relocation due to fuel fragmentation can significantly change the heat split, further affecting the thermal performance and safety of dual-cooled annular fuels. Considering the rarity of in-pile irradiation data for annular fuels, three different fuel relocation and its recovery models were developed based on modifications of relocation models for solid fuel rods or the mechanistic relocation simulation of pre-fragmented, interacting and deformable fuel fragments. The three relocation models were implanted into the fuel performance analysis code FROBA-ANNULAR, and thermo-mechanical simulation was carried out to investigate the influence of fuel relocation on gap size variation and heat split variation in an annular fuel. The results indicate that different relocation models lead to significant difference on the prediction of gap size and heat split. Relocation models developed through mechanistic simulation of pre-fragmented fragments seems to be more reasonable, compared with the other two models developed based on unreasonable and extreme assumptions. Mechanistic simulation results indicate that radial relocation of fragmented annular fuels can be bidirectional even at the initial burnup stages, resulting in the reduction of gap sizes of both internal and external fuel-cladding gaps. In addition, it was found that the gap size variation due to radial fuel relocation significantly changed the heat split and matching-rate between heat split and flow flux distribution, exerting an influence on fuel temperature and MDNBR.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"4 3","pages":"Pages 164-170"},"PeriodicalIF":0.0,"publicationDate":"2022-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605022000345/pdfft?md5=bdc868037b0a1bc56863c1ff5ed771ea&pid=1-s2.0-S2468605022000345-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77861017","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