International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第7页

Research on applicability of containment test data for HPR1000 based on phenomena scaling 基于现象标度的HPR1000围堵试验数据适用性研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-04-28 DOI: 10.1016/j.jandt.2025.04.001

Liu Yusheng , Qiao Xuedong , Tan Sichao , Jing Jianping , Wang Ting , Zhu Yuanbing

{"title":"Research on applicability of containment test data for HPR1000 based on phenomena scaling","authors":"Liu Yusheng , Qiao Xuedong , Tan Sichao , Jing Jianping , Wang Ting , Zhu Yuanbing","doi":"10.1016/j.jandt.2025.04.001","DOIUrl":"10.1016/j.jandt.2025.04.001","url":null,"abstract":"<div><div>The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests. To address the applicability of small-scale containment test data in validation process of the containment performance analysis code, the analysis method for applicability of experimental data is proposed and developed on the basis of similarity analysis of the pressure response process in the containment. The applicability study of the test data, which are produced by some scaled containment facilities such as the HDR, Battelle and CVTR, is carried out in combination with the test parameters. The applicability of each test case isobtained respectively when they are applied to validate the containment code in case of the Large Break Loss of Coolant Accident (LBLOCA) and Main Steam Line Break Accident (MSLB) of HPR1000 nuclear power plant. The results show that the similarity criteria for pressure response process and key phenomena within the containment vessel under accident conditions can be employed to analyze the applicability of different containment tests to specified power plant. The test dataoriginating fromHDR ISP-16 and HDR ISP-23can represent the pressure transient process, results of coupling phenomena such as mass and energy release at the break, condensation near the containment shell and internals, within the HPR1000 containment in case of LBLOCA or MSLB. The distortion is either within the acceptable range or conservative for design limits of containment pressure, so that the small-scale containment test data could be suitable for verification and validation of the HPR1000 containment thermal hydraulic response analysis code.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 120-127"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144189919","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 study on data analysis utilizing the K-edge absorption spectroscopy in mixed thorium and uranium solutions 钍铀混合溶液中k边吸收光谱数据分析研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-04-29 DOI: 10.1016/j.jandt.2025.04.009

Han Yang , Li Li , Quanwei Liu , Yingjie Ma , Jianbin Zhou , Liudong Hou , Yongquan Qin , Yantao Hu , Jing Ma

{"title":"A study on data analysis utilizing the K-edge absorption spectroscopy in mixed thorium and uranium solutions","authors":"Han Yang , Li Li , Quanwei Liu , Yingjie Ma , Jianbin Zhou , Liudong Hou , Yongquan Qin , Yantao Hu , Jing Ma","doi":"10.1016/j.jandt.2025.04.009","DOIUrl":"10.1016/j.jandt.2025.04.009","url":null,"abstract":"<div><div>X-ray absorption spectroscopy stands as a non-destructive assay technique commonly employed in nuclear fuel reprocessing plants to analyze the concentration of elements such as uranium, plutonium,thorium.The processing of absorption spectroscopy directly affects the measurement outcomes. In this study, an experimental setup for <em>K</em>-edge absorption spectrum has been established. Acknowledging the constraints of conventional <em>K</em>-edge absorption spectrum analysis, two improved methods are proposed, including the optimized method, and the cumulative counting method. A mixed sample of thorium and uranium was utilized to conduct an experimental study. Upon analyzing and experimental results, it was determined that the cumulative counting method offers better accuracy and simplicity in processing the absorption spectrum. During the spectrum processing, adopting a wide region of interest enhances the precision of results and reduce relative deviations, while omitting background subtraction aids in simplifying the process. The cumulative counting method can also be extended to the <em>L</em>-edge absorption spectrum.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 156-165"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144196414","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 general two-group constants estimator for 17x17 PWR assembly configurations using artificial neural networks 基于人工神经网络的17x17压水堆组件组态双组常数估计

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-04-30 DOI: 10.1016/j.jandt.2025.04.014

Gökhan Pediz , M. Alim Kırışık

{"title":"A general two-group constants estimator for 17x17 PWR assembly configurations using artificial neural networks","authors":"Gökhan Pediz , M. Alim Kırışık","doi":"10.1016/j.jandt.2025.04.014","DOIUrl":"10.1016/j.jandt.2025.04.014","url":null,"abstract":"<div><div>In this study, a preliminary general two-group constants predictor using artificial neural networks (ANNs) for pressurized water reactor (PWR) based assembly designs is established. Users can input arbitrary assembly specifications to the trained ANN, enabling the instant generation of group constants while avoiding the high computational cost of neutron transport calculations. The input parameters encompass diverse geometric configurations, material compositions and temperatures, and burnup states. The TensorFlow platform embedded in Keras API has been used to train ANNs. The number of layers and hyperparameters used in the ANN has been determined using the KerasTuner optimization framework employing the Bayesian optimization algorithm. Serpent code has been used to generate two-group constants for random assembly configurations to obtain training and test data. The prediction measures show that ANNs can consistently estimate two-group constants for the test data assemblies. The accuracy of the trained ANN is evaluated by multiplication factor calculations for selected benchmarks. Comparisons with the reference results show that ANNs can reflect the group constants for various states of a fuel assembly with satisfactory agreement.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 90-99"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144130936","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

Validation and sensitivity analysis of the FLEXPART model using the Kincaid experiment data 使用Kincaid实验数据对FLEXPART模型进行验证和灵敏度分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-07 DOI: 10.1016/j.jandt.2025.05.002

Huanting Li , Li Yang , Yinglu Cao , Xinpeng Li , Sheng Fang

{"title":"Validation and sensitivity analysis of the FLEXPART model using the Kincaid experiment data","authors":"Huanting Li , Li Yang , Yinglu Cao , Xinpeng Li , Sheng Fang","doi":"10.1016/j.jandt.2025.05.002","DOIUrl":"10.1016/j.jandt.2025.05.002","url":null,"abstract":"<div><div>FLEXPART, a Lagrangian particle dispersion model, is widely used in atmospheric dispersion and nuclear emergency response studies. However, a systematic sensitivity analysis of this model is still lacking. This study, which utilizes meteorological data output from the WRF model (with 1 km spatial resolution and 1 h temporal resolution) to drive the FLEXPART model, validates FLEXPART using the Kincaid experiment data and assesses its sensitivity to key parameters: particle number, lower bound of the turbulence intensity, and grid size. The simulation results demonstrate that the model adequately reproduces the plume dispersion pattern and covers most monitoring stations. Furthermore, the numerical results show high consistency with the observational data (FAC2: 0.50; FB: −0.22; NMSE: 1.01). The sensitivity results indicate that the particle number and the grid size significantly influence the model performance. It is recommended to use a particle number of 1000 particles/min, a lower bound of the turbulence intensity of (0.6, 0.6, 0.3) m/s, a horizontal grid size of 100 m, and a vertical grid size of 10 m for similar scenarios. These findings offer actionable guidance for parameter selection in nuclear emergency scenarios and similar dispersion studies, supporting improved model reliability and applicability.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 110-119"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144168158","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

Characteristics of aerosol release from fuel cladding breakage in lead-bismuth reactors 铅铋堆燃料包壳破裂后气溶胶释放特性研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-08 DOI: 10.1016/j.jandt.2025.05.001

Yuchen Li , Yanmin Zhou , Haifeng Gu , Yichen Zhang , Pingting Jiang , Zhenghui Wang

{"title":"Characteristics of aerosol release from fuel cladding breakage in lead-bismuth reactors","authors":"Yuchen Li , Yanmin Zhou , Haifeng Gu , Yichen Zhang , Pingting Jiang , Zhenghui Wang","doi":"10.1016/j.jandt.2025.05.001","DOIUrl":"10.1016/j.jandt.2025.05.001","url":null,"abstract":"<div><div>Lead-bismuth reactors are the most promising type among the fourth-generation reactors. When the reactor operates for a long time, the fuel cladding may develop small breaches due to high temperatures and other factors, leading to the release of radioactive fission products into the coolant, which increases the radioactivity in the primary circuit. Research on the release of fission products from damaged fuel cladding in lead-bismuth reactors is important, as the findings can provide theoretical support and model foundations for source term assessment and safety system design in severe accidents involving lead-bismuth fast reactors. This paper conducts experimental studies on the release of fission gases and aerosols from damaged fuel cladding in lead-bismuth reactors, exploring the impact of different breach sizes on the release characteristics of fission gases and aerosols. The study has obtained transient release patterns of fission gases and aerosols under various conditions. By calculating the total release fraction of aerosols, the impact of breach size on aerosol removal has been assessed. Research results indicate that when a small breach occurs in the fuel cladding, the release of fission gases follows an exponential decay pattern. The release rate of iodine within the fuel cladding is consistent with the release rate of mixed fission gases. As the release progresses, the gases will be completely released to the coolant. The release characteristics of aerosols is different from those of gases; aerosols are affected by sedimentation effects, which prevent complete release. As the standing time increases, the amount of sedimentation also increases.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 81-89"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144130935","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

Experimental progress of 209Bi(n,G)210gBi cross-section measurement based on polonium self-deposition technique 基于钋自沉积技术的209Bi(n，G)210gBi截面测量实验进展

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-28 DOI: 10.1016/j.jandt.2025.05.008

Feng Xu , Yushou Song , Size Chen , Jinlong wang , Guobing Yu , Deyun Wen , Dan Xiao , Yong Guo , Lei Ren

{"title":"Experimental progress of 209Bi(n,G)210gBi cross-section measurement based on polonium self-deposition technique","authors":"Feng Xu , Yushou Song , Size Chen , Jinlong wang , Guobing Yu , Deyun Wen , Dan Xiao , Yong Guo , Lei Ren","doi":"10.1016/j.jandt.2025.05.008","DOIUrl":"10.1016/j.jandt.2025.05.008","url":null,"abstract":"<div><div>Lead bismuth eutectic alloy (LBE:44.5 wt%Pb+55.5 wt%Bi) is an important coolant candidate for fourth-generation reactors due to its good physical and chemical properties. However, the risk of radioactive polonium radiation protection caused by liquid lead-bismuth alloy as the main coolant of the reactor has become an important problem that must be paid attention to in the development of the lead-bismuth reactor. The <sup>209</sup>Bi(n,γ)<sup>210g</sup>Bi reaction is a key source term for <sup>210</sup>Po generation in lead-bismuth reactors. Accurate measurement of the <sup>209</sup>Bi(n,γ)<sup>210g</sup>Bi cross section is important for assessing the risk of radioactive polonium release from lead-bismuth reactors. At present, the reaction cross section can be measured by the online method of directly measuring the cascade prompt gamma rays of the capture reaction or the offline method of measuring <sup>210g</sup>Bi decay product <sup>210</sup>Po. To solve this problem, a<sup>210</sup>Po self-plating enrichment method based on isotope dilution α energy spectrometry with electrode movement is proposed. By optimizing the key conditions of the self-plating process, the enrichment rate of <sup>210</sup>Po is effectively improved, which will help solve the problem of insufficient sensitivity of the direct measurement method of <sup>209</sup>Bi(n,γ)<sup>210g</sup>Bi low yield. The experimental results show that this technique is suitable for <sup>209</sup>Bi(n,γ)<sup>210g</sup>Bi post-processing measurement, and is of great significance for the study of neutron irradiation products of lead-bismuth eutectic alloys.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 139-146"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144189797","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

Thermal mechanical response analysis of the first wall in laser initial confinement fusion reactors 激光初始约束聚变反应堆第一壁热力学响应分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-15 DOI: 10.1016/j.jandt.2025.05.004

Xinze Li , Bingqian Zhang, Chengcheng Tong, Ronghua Chen, Dalin Zhang, Wenxi Tian, Suizheng Qiu

{"title":"Thermal mechanical response analysis of the first wall in laser initial confinement fusion reactors","authors":"Xinze Li , Bingqian Zhang, Chengcheng Tong, Ronghua Chen, Dalin Zhang, Wenxi Tian, Suizheng Qiu","doi":"10.1016/j.jandt.2025.05.004","DOIUrl":"10.1016/j.jandt.2025.05.004","url":null,"abstract":"<div><div>This paper investigates the thermo-mechanical response characteristics of the first wall of a laser initial confinement fusion reactor under periodic transient heat flux loads. The first wall of the reactor experiences extremely high instantaneous heat loads during the target compression process, leading to temperature and thermal stress oscillations, which can cause material fatigue damage. Finite element analysis is employed to model the first wall under various chamber sizes and target energies, and the thermal-mechanical response characteristics of W, W/Cu functional gradient materials, and blanket materials are studied under thermal shock and cyclic loading conditions. The results indicate that W/Cu functional gradient materials can effectively alleviate thermal stresses on the first wall, especially in addressing the stress concentration at the interface. Fatigue life of the first wall is assessed using the Basquin model, and it is found that increasing the chamber size and increasing the target frequency can significantly reduce temperature oscillations and thermal stresses, thus extending the service life of the first wall. This study provides valuable design references for the optimization and design of the first wall in laser initial confinement fusion reactor.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 147-155"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144196439","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

Numerical simulation study on oxygen transport in the natural circulation process of liquid lead–bismuth eutectic 液态铅铋共晶自然循环过程中氧输运的数值模拟研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-09 DOI: 10.1016/j.jandt.2025.04.008

Xiangdong Wang , Wentao Guo , Shengfei Wang , Haicai Lv , Fang Liu , Zhangpeng Guo

{"title":"Numerical simulation study on oxygen transport in the natural circulation process of liquid lead–bismuth eutectic","authors":"Xiangdong Wang , Wentao Guo , Shengfei Wang , Haicai Lv , Fang Liu , Zhangpeng Guo","doi":"10.1016/j.jandt.2025.04.008","DOIUrl":"10.1016/j.jandt.2025.04.008","url":null,"abstract":"<div><div>The lead-cooled fast reactor uses liquid lead–bismuth eutectic (LBE) as the coolant. In accident conditions, a passive cooling system is used to remove the heat from the reactor core to avoid meltdown. Therefore, the natural circulation ability of liquid LBE directly affects the inherent safety of the lead-cooled fast reactor. Due to the certain corrosiveness of LBE to the metal pipe wall, oxygen control technology needs to be used to avoid corrosion. So it is of great significance to understand the transport characteristics of oxygen concentration in LBE with natural circulation. In order to study the oxygen transport characteristics of LBE in natural circulation circuit, this article uses FLUENT software to conduct CFD simulation using the oxygen transport model. The mass flow rate of natural circulation under different power conditions in the cold and hot sections was discussed. The effects of different LBE temperatures and different inlet velocities of the expansion tank on oxygen transport were discussed in a classified manner. Based on this, the simulated values were fitted to obtain the Sh number relationship (oxygen transport relationship) for different flow patterns in the expansion tank under natural circulation conditions. The simulation results can provide a reference for the design and research of lead-cooled fast reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 166-176"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144196415","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

Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident NuScale iPWR被动安全壳在失冷剂事故中冷却行为的热水力研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-06-01 Epub Date: 2025-05-13 DOI: 10.1016/j.jandt.2025.05.003

Zhexi Guo, Sicong Xiao

{"title":"Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident","authors":"Zhexi Guo, Sicong Xiao","doi":"10.1016/j.jandt.2025.05.003","DOIUrl":"10.1016/j.jandt.2025.05.003","url":null,"abstract":"<div><div>Operation of the NuScale emergency core cooling system (ECCS) was studied, with an emphasis on the crucial role of the containment vessel (CNV). Design parameters such as vessel volume were varied to explore the resulting thermal hydraulic (TH) behaviour at the system level in the case of a loss-of-coolant accident (LOCA) initiated by a chemical and volume control system (CVCS) injection line break within the CNV. Results showed that an important phenomenon of concern was core fluid vaporisation, which could be suppressed with smaller vessels to keep system pressure and water levels in the core higher. Break elevation was found to be less significant, unless it were to occur below the reactor recirculation valves (RRV), which would result in backflow into the core and disrupt natural circulation. Increasing nominal power from 160 MW to 250 MW did not result in very different TH phenomena and could be managed by similar CNV designs and properties. Combined effects of a break at high elevation at the higher power of 250 MW appeared to reduce overall in-core vaporisation. Long-term cooling of the system was finally shown to be ensured up to 3 days with monotonously decreasing or constant core void fractions and pressures.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 2","pages":"Pages 100-109"},"PeriodicalIF":0.0,"publicationDate":"2025-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138716","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

Drop analysis and numerical simulation of spent fuel transfer basket 乏燃料输送篮跌落分析及数值模拟

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-03-01 Epub Date: 2025-04-29 DOI: 10.1016/j.jandt.2025.04.012

Jiaojiao Dong, Chen Wang, Xiaogang Xu, Siqi Zhu

{"title":"Drop analysis and numerical simulation of spent fuel transfer basket","authors":"Jiaojiao Dong, Chen Wang, Xiaogang Xu, Siqi Zhu","doi":"10.1016/j.jandt.2025.04.012","DOIUrl":"10.1016/j.jandt.2025.04.012","url":null,"abstract":"<div><div>The spent fuel transfer basket is responsible for the transfer and storage of the assemblies, and the fall condition of the basket needs to be considered to ensure the critical safety of the spent fuel assemblies during the operation process. According to the actual process, two working conditions are assumed, i.e. vertical fall of the basket at a height of 1 m from the bottom of the pool in the air condition and tipping over underwater. The key parameters in the numerical simulation are verified by carrying out experiments to obtain numerical simulation results with high accuracy. Aiming at the tipping process which is greatly influenced by the fluid resistance, the underwater tipping test of the 1:2 model of the basket is carried out to obtain the influence of the fluid on the tipping of the basket and to obtain the equivalent fall angle of the tipping fall in the air of the basket. The equivalent stiffness of the assemblies in the model is corrected by the 1:1 model 1-m vertical drop test, and the corrected numerical simulation agrees well with the experimental results, which indicates that the drop calculation model obtained in this study can be applied to the actual design.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 1","pages":"Pages 55-62"},"PeriodicalIF":0.0,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143950464","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