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

{"title":"The formation and stripping mechanism of oxide film on stainless steel surfaces","authors":"Zhenqi Shen ,&nbsp;Guangkai Wang ,&nbsp;Fang Bao ,&nbsp;Tianhao Liu ,&nbsp;Yang Fei ,&nbsp;Zhenni Xie ,&nbsp;Yuxin Liang ,&nbsp;Xincheng Wang","doi":"10.1016/j.jandt.2024.12.004","DOIUrl":"10.1016/j.jandt.2024.12.004","url":null,"abstract":"<div><div>Typically, equipment in nuclear facilities operating in nitric acid media are made of stainless steel. Retired equipment exhibits relatively high levels of radioactivity, and chemical cleaning is a cost-effective and efficient method for decontamination. Currently, the formation rules of oxide films on stainless steel surfaces during long-term immersion in nitric acid systems, leading to the development of strong oxidizing chemical decontamination agents, were investigated. It was found that oxide film of stainless steel immersed in nitric acid solution for about 150 days stabilized at a thickness of around 6 μm. Subsequently, an inorganic acid-based strong oxidizing decontamination agent was developed with nitric acid as the main component. Meanwhile, two distinguished additives (A and B) were developed to promote the stripping of oxide film. The optimized ingredients were 3.3 mol/L nitric acid, 5 wt% additive A, and 0.12 wt% additive B, which resulted in a removal thickness of up to 27.35 μm immersed at 30 °C for 10 min. Furthermore, the stripping mechanism of oxide films on stainless steel surfaces was proposed, revealing the transition from a passivated state to an over-passivated state in nitric acid environments and leading to intergranular corrosion and potential grain detachment.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 164-170"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100592","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 the configuration scheme and system parameter optimization of pumps in the component cooling system and sea water system for HPR1000","authors":"Weiguang Zhao ,&nbsp;Pei Yu ,&nbsp;Xiaobo Zeng ,&nbsp;Guangming Fan ,&nbsp;Zhaoming Meng ,&nbsp;Changqi Yan","doi":"10.1016/j.jandt.2024.12.009","DOIUrl":"10.1016/j.jandt.2024.12.009","url":null,"abstract":"<div><div>The Component Cooling System (CCS) and the Sea Water System (SWS) of HPR1000 are heat removal systems related to the safety of nuclear power plants. In both CCS and SWS, the configuration of pumps is closely associated with system design and energy consumption. This paper addresses the issue of high energy consumption and poor economic performance of the current configuration scheme that relies on a single large-capacity constant frequency pump. Based on a mathematical model for CCS and SWS that can be utilized for optimization calculations, we propose and validate an Improved Non-dominated Sorting Genetic Algorithm II (INSGA-II) with good algorithm performance, optimizing for system design costs and total operational costs as objectives, and provide a sensitivity analysis of relevant variables. The results demonstrate that compared to the prototype values of HPR1000, the investment costs of CCS and SWS can be reduced by up to 4.65 %, and the total operational costs can be decreased by as much as 63.6 %, with the optimization effect being most significant when variable frequency pumps are used in CCS and SWS.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 2","pages":"Pages 146-157"},"PeriodicalIF":0.0,"publicationDate":"2024-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143100889","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":"Validation of transient SYSTEM analysis code GINKGO based on a SLB TEST","authors":"Si Si Wang,&nbsp;Ke Wang,&nbsp;Zhong Ying Ma,&nbsp;Ting Wang","doi":"10.1016/j.jandt.2024.05.006","DOIUrl":"10.1016/j.jandt.2024.05.006","url":null,"abstract":"<div><p>Transient System analysis code is an important tool for reactor thermal hydraulic analysis. GINKGO is a transient system analysis in-house code developed by China General Nuclear Power Corporation (CGN). The code is primarily used for the analysis of Non-Loss of Coolant Accident (Non-LOCA)transients. Steam Line Break (SLB) accident is one of the representative accident condition of Non-LOCA transients. To validate the capability of GINKGO to simulate SLB accident scenario, a test of small non-isolatable main steam line break (10 %) at hot standby conditions, PKL III B5.1 test is selected. The characteristic of PKL facility, the PKL III B5.1 test sequence and the validation process of PKL III B5.1 test by GINKGO code are briefly introduced in the paper. In validation process, GINKGO simulates the steady state first and then simulates the transient process by setting reasonable initial conditions and boundary conditions. Then, the calculation results of key parameters including temperature in hot leg, temperature in cold leg, Steam Generator (SG) secondary side pressure and Pressurizer (PZR) pressure are compared with the experimental results. Through the rationality of the calculation results, it shows that GINKGO has the capability to simulate the key phenomena in a typical SLB accident scenario.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 1-8"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000176/pdfft?md5=092f88272be9b36b50ccdd6e9411735f&pid=1-s2.0-S2468605024000176-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141230928","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":"Advancements in the research of high-temperature gas-cooled reactor fuel via additive manufacturing techniques","authors":"Haoyu Liao,&nbsp;Ting Zhang,&nbsp;Chenxi Li,&nbsp;Changbin Tang,&nbsp;Yuanming Li,&nbsp;Ping Chen","doi":"10.1016/j.jandt.2024.05.005","DOIUrl":"10.1016/j.jandt.2024.05.005","url":null,"abstract":"<div><p>Recent advances in high-temperature gas-cooled reactor fuel research have captured the global spotlight, as modern operating contexts demand innovative responses to unprecedented challenges. This article offers an overview of contemporary directions in high-temperature gas-cooled reactor fuel design worldwide, pinpointing the shared operational demands and development patterns characteristic of these advanced fuels. It presents an innovative advanced fuel design, which incorporates multi-layer coated particles to boost safety and employs high-temperature-resistant silicon carbide (SiC) ceramics, seamlessly integrated with cutting-edge additive manufacturing (AM) techniques. This design heralds a shift toward modular fuel element solutions that are easily scalable, supporting the diverse array of advanced reactors and marking a significant step forward in the quest for high-temperature gas-cooled reactor fuel technologies.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 14-20"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000164/pdfft?md5=0a1c200bb294edb32495b830e4588c61&pid=1-s2.0-S2468605024000164-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141145487","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":"Investigation on crack propagation of SiCf/SiC composite via PF-CZM method","authors":"Bowen Qiu ,&nbsp;Zhuang Li ,&nbsp;Shuang Liang ,&nbsp;Cheng Zhang ,&nbsp;Chong Wei ,&nbsp;Zhong Xiao","doi":"10.1016/j.jandt.2024.06.001","DOIUrl":"10.1016/j.jandt.2024.06.001","url":null,"abstract":"<div><p>SiCf/SiC composite materials had become a research hotspot in the fields of aerospace materials and nuclear materials in recent years. This materials have high strength, radiation resistance, high temperature resistance and oxidation resistance. However, SiC is a typical brittle material prone to being damaged due to stress concentration in some cases, leading to cracks and even component failure. Therefore, this study employed an accurate and efficient numerical simulation method, phase field cohesive zone method (PFCZM), to investigate the microscopic damage behavior of SiCf/SiC composite materials. This study aimed to illustrate the influence of the thickness of pyrolytic carbon (PyC) interface layer on the failure behavior of SiCf/SiC composite materials. Firstly, this study conducted relevant verification of PFCZM. Secondly, numerical simulations of crack initiation, crack propagation, and final damage were performed by establishing a representative volume element (RVE). Observing the simulation results, it was noted that at low strain, when PyC was thicker, the SiCf/SiC composite materials exhibited slight damage to both the SiC matrix and PyC interface. Conversely, when PyC was thinner, the SiCf/SiC composite materials displayed only slight damage to PyC interface. As the PyC thickness increases, the damage to PyC interface decreases, while the damage to SiC becomes more prominent. Simultaneously, a competitive relationship in the damage behavior of SiCf/SiC composite materials can be found after the SiC matrix has cracked: When PyC is thinner, the damage primarily manifests as interlayer delamination between the SiC matrix and the SiC fiber. Conversely, when PyC is thicker, it manifests in PyC failure. Finally, this study concluded that, in this investigation, a PyC interface layer with a thickness of 0.1 μm exhibited the most effective protective effect on SiCf/SiC composite material.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 9-13"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000255/pdfft?md5=f7543088b48cbab3ea1264c1033efb4b&pid=1-s2.0-S2468605024000255-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141410939","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":"Investigation on numerical simulation models for heat transfer characteristics of lead-bismuth eutectic","authors":"Zihan Qin,&nbsp;Ling Chen,&nbsp;Yongfa Zhang,&nbsp;Hao Qin","doi":"10.1016/j.jandt.2024.09.002","DOIUrl":"10.1016/j.jandt.2024.09.002","url":null,"abstract":"<div><div>The liquid lead-bismuth eutectic (LBE) has distinct thermal and physical properties that challenge the effectiveness of traditional RANS turbulence models and turbulent Prandtl number (<em>Pr</em>_t) models for accurately simulating its flow and heat transfer characteristics. This study aims to investigate numerical simulation models of LBE based on existing experimental data. Simulations are conducted using different combinations of three turbulence models and four <em>Pr</em>_t models, and comparisons are made with experimental data. The local and overall heat transfer characteristics of LBE are analyzed, and the applicability of each model combination is evaluated. Results indicate that for simulating local heat transfer characteristics, the SST <em>k</em>-<em>ω</em> turbulence model combined with the <em>Pr</em>_t model proposed by Cheng et al. yields the highest accuracy. Additionally, the empirical correlation for heat transfer proposed by Kutateladze provides the best prediction of the local Nusselt number. For overall heat transfer characteristics, the combination of the SST <em>k</em>-<em>ω</em> turbulence model and the <em>Pr</em>_t model introduced by Reynolds et al. demonstrates the highest accuracy and applicability. This investigation might offer a pivotal benchmark for the discernment of appropriate computational fluid dynamics (CFD) models for liquid metal breeder reactor (LMBR) applications utilizing lead-bismuth eutectic (LBE) as coolant.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 57-66"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000292/pdfft?md5=5146a2a687a8a5c8d56d28a654eca07f&pid=1-s2.0-S2468605024000292-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142316135","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 on thermal-hydraulic performance of helically coiled tubes steam generator of lead-bismuth reactor","authors":"Dandan Gu,&nbsp;Lina Zhu,&nbsp;Changzhi Xiao,&nbsp;Hongyi Yang","doi":"10.1016/j.jandt.2024.07.002","DOIUrl":"10.1016/j.jandt.2024.07.002","url":null,"abstract":"<div><p>The advantages of the helically coiled tube steam generator lie not only in its high heat transfer efficiency but also in its compact structure, which is initially used in Pb–Bi–SCO₂ reactors, and its main function is to transfer the heat from the lead-bismuth side to the SCO₂ side. In this experiment, two sets of helical heat exchanger tubes are selected as the objects of investigation. Experiments are conducted on the heat transfer performance of a helical tube to study the influence of structural parameters, mass flux, pressure, and heat flux on the heat transfer characteristics during subcooled boiling and post-dryout. Considering the results of the study, new heat transfer characteristic correlations are derived for the subcooled boiling and post-dryout heat transfer regions, which provide data and theoretical support for the engineering application and thermal hydraulic analysis of steam generators.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 32-42"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000279/pdfft?md5=0871830ece83919aa5704f5de3ac1a35&pid=1-s2.0-S2468605024000279-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141850612","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":"Multi-objective optimization research of open and closed air brayton cycle","authors":"Meihui Song ,&nbsp;Yiran Qian ,&nbsp;Yuyang Leng ,&nbsp;Tao Liu ,&nbsp;Lianyun Yu ,&nbsp;Weixiong Chen","doi":"10.1016/j.jandt.2024.07.001","DOIUrl":"10.1016/j.jandt.2024.07.001","url":null,"abstract":"<div><p>Air Brayton cycle obtains the significant advantages of high efficiency, compact structure, easy medium requirement and so on, which is one of the most suitable choices for the mobile nuclear power conversion system. In this paper, system-component combined design method is used to establish the open and closed air Brayton cycle of diverse configurations. Based on performance and compactness targets, the multi-objective optimization is carried out to find the optimal design by nondominated sorting genetic algorithm II and entropy weight method. According to the analysis, the closed intercooling reheating recuperating cycle is the configuration with the best comprehensive performance, with cycle efficiency of 33.58 %, power density of 175.39 kW m⁻³ and power mass ratio of 68.60 kW t⁻¹. The open simple recuperating cycle is the configuration with the smallest volume and mass, with cycle efficiency of 20.01 %, system volume of 6.56 m³ and system mass of 19.98 t. The closed intercooling recuperating cycle is the most balanced configuration, with cycle efficiency of 31.33 %, volume of 9.06 m³ and system mass of 24.97 t. Based on the optimal results, off-design performance analysis of different environmental conditions is also carried out for the configurations above.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 21-31"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000267/pdfft?md5=8c691f87b18e51d0f5eba98a337cee82&pid=1-s2.0-S2468605024000267-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141637275","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":"The main structural-phase states of interaction between model corium of a nuclear reactor and a sacrificial material based on Al2O3 and a lead layer","authors":"M.K. Skakov ,&nbsp;V.V. Baklanov ,&nbsp;I.M. Kukushkin ,&nbsp;M.K. Bekmuldin ,&nbsp;A.S. Akaev ,&nbsp;M.N. Azbergenov ,&nbsp;A.V. Gradoboev","doi":"10.1016/j.jandt.2024.09.001","DOIUrl":"10.1016/j.jandt.2024.09.001","url":null,"abstract":"<div><p>This article presents the results of post-experimental studies of the interaction of model light water reactor corium with a candidate sacrificial material (SM) based on Al₂O₃ (corundum) and a lead layer. The purpose of the experiments was to study the features of the interaction of the sacrificial material with corium in the conditions of simulating a severe accident with core meltdown. An active interaction of corium with the sacrificial material was established as a result of the research. This is confirmed by the fact that part of the sacrificial material entered into physical and chemical interaction with the components of corium with the formation of liquid-phase reaction products and, as a result, creating different conditions for the crystallization of corium. The study of the phase composition showed that the microstructure of solidified corium in various areas after interaction with the sacrificial material is represented by a phase of the (U,Zr, …)O_2±x type with different types of crystal lattice. At the same time, the phase analysis showed a close correspondence of the composition of the solid solutions to the initial ratio of uranium and zirconium in the model corium charge. This allows us to conclude that liquid zirconium interacted with aluminum oxide during the experiment. Thus, it was concluded that the proposed sacrificial material is promising based on the results obtained and the identified features of the interaction of aluminum oxide with the lead layer and the corium of a nuclear reactor.</p></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"6 1","pages":"Pages 43-56"},"PeriodicalIF":0.0,"publicationDate":"2024-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S2468605024000280/pdfft?md5=ff9269b14ac7c7a950a929cf7f5b0564&pid=1-s2.0-S2468605024000280-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142173650","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 of aerosol removal by passive containment cooling system","authors":"Pingwen Ou,&nbsp;Peng Chen","doi":"10.1016/j.jandt.2024.05.004","DOIUrl":"10.1016/j.jandt.2024.05.004","url":null,"abstract":"<div><div>Passive containment cooling system (PCCS) can effectively remove the heat from the containment and maintain the pressure and temperature in case of accident, thus ensuring the containment integrity. PCCS introduces thermophoresis and diffusiophoresis aerosol deposition and changes the distribution of fission products in the containment, thus affecting containment dose evaluation and off-site source term assessment. In this paper, a Generation III PWR nuclear power plant model with PCCS is established by using ASTEC (Accident Source Term Evaluation Code), and the removal effect of PCCS on fission product aerosol is studied under the large break loss of coolant accident. The results show that one train of PCCS can increase the aerosol removal factor for approximately 0.05/h. Taking five trains of PCCS in this study as an example, aerosol deposition effect introduced by PCCS reduces suspended aerosol in containment by 10 %–20 %, and the effect is more significant in the PCCS-locating compartment. Meanwhile, the gas flow among compartments changes the aerosol mass in the adjacency of the PCCS-locating compartment. This phenomenon weakens the removal effect in the PCCS-locating compartment and make the aerosol concentration in adjacent compartments tend to be consistent.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"5 4","pages":"Pages 200-206"},"PeriodicalIF":0.0,"publicationDate":"2023-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141133948","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}