International Journal of Advanced Nuclear Reactor Design and Technology最新文献

{"title":"Investigate of electrochemical properties of Nd(II) and Nd(III) in eutectic LiCl–KCl molten salt","authors":"Xinyu Zhang,&nbsp;Wentao Zhou,&nbsp;Dezhong Wang","doi":"10.1016/j.jandt.2024.10.002","DOIUrl":"10.1016/j.jandt.2024.10.002","url":null,"abstract":"<div><div>Pyroprocessing is one of the promising methods to reprocess the spent nuclear fuel. Neodymium (Nd) constitutes about one third of rare earth in fission products, however, the electrochemical properties of Nd have not been thoroughly understood, especially the exchange current. In the present study, the transit electrochemical methods of cyclic voltammetry and square wave voltammetry were applied to study the properties of Nd in LiCl–KCl system. The reduction of Nd³⁺ was determined as a two-step reaction with Nd³⁺/Nd²⁺ to be a reversible but Nd²⁺/Nd to be a quasi or irreversible reactions. Additionally, parameters such as the diffusion coefficient, exchange current, and apparent potential were derived from the cyclic voltammetry curves. Their relationship with temperature and concentration was investigated under 723 K, 773 K, 823 K with concentrations of 3 wt%, 6 wt% and 9 wt%. The present study contributes significantly to rich the database of Nd in LiCl–KCl molten salt and to understand the electrochemical behaviors of Nd during electrochemical separation of spent nuclear fuel.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 90-98"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661127","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}

{"title":"Method of out-of-pile high-temperature tests of low-melting materials in conditions of modeling a severe nuclear reactor accident","authors":"Kuanyshbek Toleubekov ,&nbsp;Mazhyn Skakov ,&nbsp;Viktor Baklanov ,&nbsp;Maxat Bekmuldin ,&nbsp;Assan Akaev","doi":"10.1016/j.jandt.2024.10.003","DOIUrl":"10.1016/j.jandt.2024.10.003","url":null,"abstract":"<div><div>This article is devoted to the technique for conducting experiments to study the nature of the interaction of low-melting metals with corium in the conditions of simulating a severe nuclear reactor accident. A feature of such experiments is the need to ensure thermophysical conditions for the occurrence of a severe accident and the organization of the interaction of the metals under study with liquid corium after its obtaining. In this regard, special device for solid metal fragments discharging into liquid corium were designed and manufactured to provide such conditions. At the same time, modeling of the temperature field of the experimental assembly of the VCG-135 test bench to study the interaction between model corium and candidate metal-coolers in the conditions of a severe accident was conducted. The need for modeling is associated with to the probability of metal melting in the discharge device due to the heat flow from the heating crucible of the experimental assembly. The operability of the developed technique was shown as a result of modeling and subsequent experiments at the VCG-135 test bench. The experiments made it possible to obtain products of the interaction of metals with corium for further study of their structural and phase state.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 99-107"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661087","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}

{"title":"Analysis of flow distribution and bypass flow for DHR-400 low temperature heating reactor","authors":"Yan Zhang ,&nbsp;Mingyu Liu ,&nbsp;Jing Chen ,&nbsp;Shizhao Song ,&nbsp;Hao Yin","doi":"10.1016/j.jandt.2024.12.003","DOIUrl":"10.1016/j.jandt.2024.12.003","url":null,"abstract":"<div><div>Flow distribution at the core inlet and the fraction of bypass flow are important aspects of thermal hydraulic design for reactor. Three-dimensional fluid dynamics analysis was used to numerically simulate the overall flow phenomenon in the pool type low temperature heating reactor. Model of porous media was used to modify fuel assemblies. Thermal hydraulic parameters such as flow distribution of inlet and bypass flows were obtained. The results show that the deviation between the maximum flow rate and the average flow rate at the core inlet is 4.36 %, and the deviation between the minimum flow rate and the average flow rate is −3.08 %. The fraction of bypass flow of reflectors is 2.3 %, and the fraction of bypass flow between baffles and core shroud is 0.34 %. The total pressure drop from inlet to outlet is 56.8 kPa. It can be concluded that the distribution of inlet is relatively uniform, and the fraction of bypass flows is within the allowable range. The calculation results can provide a basis for subsequent engineering design.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 131-138"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100891","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}

{"title":"A comparative neutronic analysis of U–Th fuels as potential alternatives to traditional UO2 fuel for enhanced performance and safety in the ACP-100 SMR","authors":"Md. Abidur Rahman Ishraq ,&nbsp;Sabyasachi Roy","doi":"10.1016/j.jandt.2025.02.002","DOIUrl":"10.1016/j.jandt.2025.02.002","url":null,"abstract":"<div><div>This study examines U–Th-based alternative fuel options for the ACP-100 reactor core through a neutronic lens. Neutronic simulations were performed utilizing the Monte Carlo SERPENT code for 540 EFPDs. This study seeks to overcome the limitations of traditional UO₂ fuel by exploring alternatives that provide enhanced performance, greater resource availability, and increased safety. This study evaluates U–Th-based carbide, nitride, and oxide fuels in comparison to conventional UO₂. Among the alternatives, U–Th-based fuels exhibit a 40 % higher conversion ratio compared to UO2, along with superior neutron flux and enhanced transmutation capabilities with isotopes relative to other evaluated fuel elements. A suppressed trend is noted in <em>keff</em> and burnup for U–Th-based fuels in comparison to UO₂. Nonetheless, U–Th fuels demonstrate favorable outcomes regarding safety parameters, including ppfs, MTC, FTC, and βeff. The U–Th-based fuel elements with added nitride and carbide exhibit reductions in ppfs of 1.78 % and 1.5 %, respectively, when compared to UO2. The findings offer important insights into the neutronic behavior of U–Th-based fuels and their potential to enhance the performance and safety of the ACP-100 reactor.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 177-184"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420535","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}

{"title":"Experimental study of aerosol diffusiophoresis deposition in steel containment under external cooling measures","authors":"Hong Ding,&nbsp;Peizheng Hu,&nbsp;Wenbin Zou,&nbsp;Zhizhou Zhu,&nbsp;Lili Tong,&nbsp;Xuewu Cao","doi":"10.1016/j.jandt.2024.12.008","DOIUrl":"10.1016/j.jandt.2024.12.008","url":null,"abstract":"<div><div>The steel containment of an advanced passive Pressurized Water Reactor adopted a passive containment cooling system to transfer residual heat to the environment through the steel containment, where aerosol is deposited with the steam condensation inside the containment. This paper focuses on aerosol removal with strong steam condensation by external cooling of steel containment. The experiments on aerosol natural deposition at 500 kPa (a) with different conditions were conducted, including steady-state aerosol gravity deposition, thermophoresis deposition with non-condensable ambient gas, and diffusiophoresis deposition under the condition of 75 % steam volume fraction. The deposition characteristics of gravity, thermophoresis, and diffusiophoresis were analyzed using the cumulative mechanism separation method. The results show that diffusiophoresis is the dominant mechanism for aerosol removal under the wall steam condensation conditions. Three different diffusiophoresis models were evaluated with the experimental results. The present models significantly underestimate the experimental results, with a 70 % error between the prediction and experiments.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 139-145"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100888","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}

{"title":"Research on reactor power prediction of nuclear power plant based on multivariate optimization GRU model","authors":"Canyi Tan ,&nbsp;Bo Wang ,&nbsp;Jiangkuan Li ,&nbsp;Jie Chen ,&nbsp;Biao Liang ,&nbsp;Shangcai Zheng ,&nbsp;Rui Han ,&nbsp;Ruifeng Tian ,&nbsp;Sichao Tan","doi":"10.1016/j.jandt.2024.10.001","DOIUrl":"10.1016/j.jandt.2024.10.001","url":null,"abstract":"<div><div>In the operation of nuclear power plants, the accurate prediction of power change trends is crucial for ensuring safety and stability. In this work, a ML-GRU-RS method, based on model-agnostic meta-learning (MAML), gate recurrent unit (GRU), and random search optimization, is proposed for the long-term prediction of key parameters of nuclear power plants. This method combines the fast adaptability of MAML, the time series data processing capability of GRU, and the optimization efficiency of random search to achieve high-precision predictions under varying power conditions. The results demonstrate that this method can effectively predict the future trends of key parameters in nuclear power plants. The ability of operators to anticipate these trends has been significantly enhanced, contributing to the overall safety of the nuclear power plants.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 78-89"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661126","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}

{"title":"Ensuring safety in the Air transport of radioactive materials: Regulations, methodologies, and probabilistic safety assessment","authors":"Manorma Kumar,&nbsp;Guillem Cortes-Rossell","doi":"10.1016/j.jandt.2025.02.001","DOIUrl":"10.1016/j.jandt.2025.02.001","url":null,"abstract":"<div><div>The paper is focused on comprehensive study on air transportation of radioactive materials, covering regulations and requirements, hazard identification and proposed a methodology to apply the Probabilistic Safety Assessment to assess the nuclear safety risks associated with the transportation of radioactive material. The systematic methodology starts with identification of different accident scenarios associated with air transportation and estimation of its likelihood that leads to consequences.</div><div>The accidents involving the radioactive material has sensitive information and such information is not publicly available. Therefore, the generic accident data can be utilized to develop the probabilistic models and to quantify the risk associated with radioactive material transportation. Even though, the probability of occurrence of an event associated with air transportation of radioactive material is low, but not negligible, hence these events can be correlated to low probability high consequences extreme events. As the associated statistics for air nuclear accidents are insufficient for the development of probabilistic models, so the generic air accident data can be used to give the predictions for accident frequency. For the data analysis a new tool is developed, and its technical bases and integration with probabilistic models are discussed in this paper. A comprehensive probabilistic assessment for air accidents is performed and an event tree for air accidents is proposed with detailed discussion on risk quantification and results insights; and finally, conclusions are presented.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 185-195"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143454799","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}

{"title":"Thermal hydraulic numerical analysis and experimental validation of GY-20A cobalt source transportation cask","authors":"Liu Jiatai,&nbsp;Geng Yiwa,&nbsp;Fang Jun","doi":"10.1016/j.jandt.2024.12.007","DOIUrl":"10.1016/j.jandt.2024.12.007","url":null,"abstract":"<div><div>Radioactive materials produced by the nuclear industry need to be properly post-processed to avoid radiation damage to people and the environment. Radioactive materials are usually transported and temporarily stored in dedicated spent fuel transportation casks. Therefore, the cask designed for transporting these radioactive materials must meet rigorous acceptance criteria and regulatory standards. The GY-20A cobalt source transportation cask is a special type cask designed for radioactive ⁶⁰Co sources transportation, with a maximum loading capacity of 200,000 Ci. In this paper, the thermal test of the transportation cask prototype is carried out, and the thermal analysis under normal conditions is analyzed based on the finite volume method code ANSYS Fluent. The predicted axial and radial temperature values are obtained. The numerical calculation results are in good agreement with the test results. The cask has complete inclusiveness under normal condition. The test and calculation results are directly used for the cask design license application.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 123-130"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100890","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}

{"title":"Analysis of explosion incidents in nuclear fuel reprocessing facilities and recommendations for their prevention","authors":"Jiaxin Liu,&nbsp;Yang Tian,&nbsp;Shuo Yang,&nbsp;Yongquan Qin,&nbsp;Xuefeng Hou,&nbsp;Yantao Hu,&nbsp;Liudong Hou,&nbsp;Jing Ma","doi":"10.1016/j.jandt.2024.12.001","DOIUrl":"10.1016/j.jandt.2024.12.001","url":null,"abstract":"<div><div>Human society endeavors to harness nuclear energy for peaceful utilization, intertwining it with global strategies to address pressing issues such as energy demand, environmental preservation, and energy infrastructure optimization. However, the proliferation of nuclear power plants has resulted in the accumulation of spent fuel, amounting a staggering 7000 tons per annum. Therefore, the safe reprocessing of spent fuel has emerged as a critical aspect of sustainable nuclear energy development. The repercussions of an explosion within a nuclear fuel reprocessing facility can be calamitous, potentially leading to radioactive emissions and triggering a nuclear incident. This review scrutinizes the lessons learned from 18 historical explosion incidents at nuclear facilities to meticulously identify their underlying causes and ramifications. We sorted the incidents based on the time of the accident and found that the generation of energetic materials in a limited space was the main cause of the explosions. Six accidents were “red oil explosions,” one of which was identified as a grade 6 nuclear accident. The use of organic solvents with chemical stability must be considered to avoid a red oil explosion. In addition, the findings emphasize the severe risk of chemical explosions during fuel reprocessing. Critical actions for ensuring the safe reprocessing of nuclear fuel include providing parameter control and equipment maintenance for evaporative units, ensuring an effective cooling system for high-level waste storage equipment, preventing the inadvertent desiccation of nitric acid substances within storage facilities, and mitigating human error through training and process design. A scrutinization of typical explosion incidents in nuclear chemical plants can support the development of viable and effective preventive measures, thereby reducing the potential risks associated with explosions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 108-116"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100893","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}

{"title":"Optimization study of lead-based reactor heat exchangers based on Dakota coupling CFD","authors":"Ge Gao,&nbsp;Chong Gao,&nbsp;Zhangpeng Guo,&nbsp;Hao Wu,&nbsp;Haicai Lyu,&nbsp;Fenglei Niu,&nbsp;Yang Liu","doi":"10.1016/j.jandt.2024.05.001","DOIUrl":"10.1016/j.jandt.2024.05.001","url":null,"abstract":"<div><div>The lead-bismuth fast reactor is one of the fourth-generation reactor types.The microchannel heat exchangers are used as intermediate heat exchangers, which significantly influence the efficiency of supercritical carbon dioxide (S–CO2) Brayton cycle power generation. It is necessary to quantify the uncertainties and optimize the design of heat transfer and pressure drop of the microchannel heat exchangers. Due to the high cost and time-consuming problem of numerical simulation, this study employs the Arbitrary polynomial chaos expansion method to generate surrogate models for simulation calculations. Specifically, the microchannel parameters of the heat exchanger, namely flow direction spacing, transverse spacing, staggered spacing, and inlet velocity, are studied for the simulation. The pressure drop and convective heat transfer coefficient of the heat exchanger are considered as the objective functions for both simulation and optimization models. Uncertainty quantification analysis, sensitivity analysis, and optimization design of those parameters are conducted. By coupling DAKOTA with CFD, an optimization objective function is established to obtain optimal design data within specified ranges.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 67-77"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660634","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}