Progress in Nuclear Energy最新文献

{"title":"Adaptive BP-PID control for steam generator level using hardware-in-the-loop simulation","authors":"YuLong Wang,&nbsp;Zhejun Sun,&nbsp;Qi Zhang,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.pnucene.2025.106203","DOIUrl":"10.1016/j.pnucene.2025.106203","url":null,"abstract":"<div><div>The steam generator, an essential component of a pressurized water reactor nuclear power plant, has level parameters that directly influence the safety and stability of the nuclear power unit. The nonlinearity of the steam generator, coupled with the ‘false level’ issue, complicates its level control. Therefore, research on steam generator level control is considered to be of great significance. Traditional level control in steam generators employs a fixed-parameter Proportional-Integral-Derivative (PID) scheme, a widely used control strategy, which can struggle to adapt to changes in operating conditions and system characteristics. By integrating the adaptive learning capability of the Backpropagation (BP) neural network into the PID framework, a BP-PID control scheme is developed, where the BP network dynamically adjusts the PID parameters in real-time based on system feedback, enabling adaptive adjustment to varying operating conditions and enhancing control effectiveness. The proposed control algorithm was tested on a hardware-in-the-loop simulation platform. Results indicate that the BP-PID control significantly outperforms the original fixed-parameter PID control, demonstrating its potential for engineering implementation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106203"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841098","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study of oxygen control model based on electrochemical oxygen pumping in liquid lead-bismuth alloys","authors":"Zhenhua Sheng,&nbsp;Weihao Wu,&nbsp;Tengjun Geng,&nbsp;Shengfei Wang,&nbsp;Fang Liu,&nbsp;Huiping Zhu,&nbsp;Haicai Lyu,&nbsp;Wentao Guo,&nbsp;Zhangpeng Guo,&nbsp;Ruixian Liang,&nbsp;Fenglei Niu","doi":"10.1016/j.pnucene.2025.106206","DOIUrl":"10.1016/j.pnucene.2025.106206","url":null,"abstract":"<div><div>To study the mass transfer characteristics of convection-enhanced oxygen transport in electrochemical oxygen pumps, a theoretical model of oxygen transport was developed based on reaction kinetics. Two-dimensional and three-dimensional mass transfer models were subsequently established using CFD software. The experimental validation showed that the model error was within 10 %. The oxygen transport process in the electrochemical oxygen pump can be broadly divided into two stages: the ion-transport-dominated mass transfer within the solid electrolyte and the diffusion-dominated mass transfer in the lead-bismuth. The oxygen enhancement rate and the minimum oxygen concentration during the oxygen reduction process were fitted to the Arrhenius equation, further confirming the accuracy of the theoretical model. Based on this validated model, flow fields were incorporated, and impeller stirring was employed to achieve convection-enhanced mass transfer, aligning closely with experimental conditions. The effects of impeller speed and temperature on oxygen transport in the lead-bismuth tank were evaluated using the multiphysics field model. The results show that convection enhancement can effectively enhance the mass transfer of electrochemical oxygen pumping in lead-bismuth without considering the limitations of solid electrolyte. By strengthening the convection effect in the lead-bismuth and the electrochemical oxygen pump, one can achieve a limited improvement in overall mass transfer efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106206"},"PeriodicalIF":3.2,"publicationDate":"2025-12-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841092","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on two-phase heat transfer enhancement in narrow rectangular channels during reflooding process","authors":"Yuanlin Yao ,&nbsp;Qianlong Zuo ,&nbsp;Penghui Zhang ,&nbsp;Jian Deng ,&nbsp;Dan Wu ,&nbsp;Jiayue Zhou ,&nbsp;Deqi Chen","doi":"10.1016/j.pnucene.2025.106207","DOIUrl":"10.1016/j.pnucene.2025.106207","url":null,"abstract":"<div><div>Reflooding serves as an essential emergency cooling strategy following reactor core exposure due to a loss of coolant accident. The compact architecture of advanced nuclear core, in contrast to conventional fuel rods, elevates the risk of localized heat flux peaks and imposes more stringent demands on reflooding heat transfer efficiency. This work experimentally investigates heat transfer enhancement methods during reflooding in narrow rectangular channels. The reflooding characteristic was firstly analyzed, then the effects of inlet parameters, nanofluids, and microstructured surfaces on the minimum film boiling temperature and cooling time were quantitatively evaluated. Experimental results demonstrate that increasing the coolant inlet velocity and subcooling significantly improve the minimum film boiling temperature and reduce the cooling time. In contrast, initial wall temperature and input heat flux exhibit negligible influence on minimum film boiling temperature but are positively correlated with cooling time. Nanofluid (Fe₃O₄ nanoparticles &lt;1 wt%) exhibit negligible heat transfer enhancement due to the vapor film preventing their surface deposition. The microstructured surface achieved the minimum film boiling temperature improvements of 4.8 %–21.1 % relative to the flat surface across the tested inlet velocities and 6.5 %–12.6 % under varying subcooling conditions, while simultaneously reducing the cooling time by 1.8 %–2.7 %.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106207"},"PeriodicalIF":3.2,"publicationDate":"2025-12-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Transformation of montmorillonite under thermal-saline conditions in HLW repositories: A systematic review","authors":"Cong Liu ,&nbsp;Yong-Gui Chen ,&nbsp;Rao-Ping Liao ,&nbsp;Zhao Sun ,&nbsp;Yu-Cheng Li ,&nbsp;Sheng-Fei Cao ,&nbsp;Wei-Min Ye","doi":"10.1016/j.pnucene.2025.106201","DOIUrl":"10.1016/j.pnucene.2025.106201","url":null,"abstract":"<div><div>The long-term stability of montmorillonite under near-field thermal and saline conditions in a high-level radioactive waste (HLW) repository is a critical concern as potential transformation of montmorillonite may occur under such circumstances. This review aims to summarize current knowledge on the transformation of montmorillonite under thermal and saline conditions relevant to HLW repositories, highlighting key mechanisms and patterns reported in the literature. Laboratory and in-situ tests indicate that montmorillonite can undergo thermo-saline transformations into illite, kaolinite, chlorite, or other minerals, and may also exhibit changes in negative layer charge. The effects of changes in negative layer charge and interlayer cations on the swelling properties of montmorillonite were analyzed. In addition to temperature and salinity, factors such as montmorillonite type, accessory minerals, pressure, reaction time, and solid/liquid ratio influence these transformations of montmorillonite.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106201"},"PeriodicalIF":3.2,"publicationDate":"2025-12-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841097","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A critical review of spent nuclear fuel drying","authors":"Ji Hwan Lim","doi":"10.1016/j.pnucene.2025.106196","DOIUrl":"10.1016/j.pnucene.2025.106196","url":null,"abstract":"<div><div>This review critically examines advanced methodologies for drying spent nuclear fuel (SNF), emphasizing the optimization of vacuum drying strategies and the role of forced helium drying (FHD). SNF drying is vital for safe storage and prevents issues like moisture-induced corrosion and hydrogen radiolysis. Traditional vacuum drying remains a staple, yet challenges like ice formation and ineffective moisture removal in damaged fuel necessitate innovations like FHD, which ensures thorough moisture elimination through active gas circulation. Empirical studies and enhanced simulations have refined understanding, showing automation and real-time moisture sensors could streamline drying, improve safety, and reduce time. Also, developments in hydride reorientation in zirconium cladding due to high temperatures during drying highlight the need for precise thermal management to prevent cladding embrittlement. Recent research emphasizes tailored drying criteria for diverse fuel types, focusing on damage-prone fuel requiring extended, heat-assisted drying strategies or supportive agents like hydrogen getters to mitigate gas build-up risks. Monitoring technologies are advancing to include dew point sensors and acoustic devices for real-time drying assessment. Thus, while vacuum methods are robust, an international collaborative effort is essential to standardize best practices for complex scenarios involving high-burnup and damaged SNF, ensuring fuel integrity and regulatory compliance across the nuclear industry.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106196"},"PeriodicalIF":3.2,"publicationDate":"2025-12-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841096","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Highly accurate and interpretable machine learning models for predicting HTC in the presence of NCG on outer vertical tubes","authors":"Huifang Zhang,&nbsp;Yapei Zhang,&nbsp;Shihao Wu,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu,&nbsp;Guanghui Su","doi":"10.1016/j.pnucene.2025.106198","DOIUrl":"10.1016/j.pnucene.2025.106198","url":null,"abstract":"<div><div>Condensation widely exists in nature and engineering. Optimizing and modeling relevant processes heavily rely on accurate heat transfer coefficient (HTC) prediction. Nevertheless, obtaining precise HTC through empirical and numerical approaches continues to present significant challenges. This study aims to develop accurate and interpretable HTC prediction models using advanced machine learning (ML) techniques. A comprehensive database of 680 experimental data points was compiled to evaluate Support Vector Regression (SVR), ensemble learning models, and the AutoGluon platform against traditional correlations. The models predicted HTC based on total pressure P_tot, the mass fraction of non-condensable gas (NCG) w_nc, the wall supercooling degree ΔT, and geometric parameters. Results demonstrate that all ML models exhibit superior performance compared to traditional correlations. Specifically, the AutoGluon platform outperforms other ML methods in predicting the HTC, according to the prediction accuracy and generalization ability comprehensively, with an MAPE of 3.16 %, RMSE of 69.46 and an R² of 0.993. Furthermore, SHapley Additive exPlanations (SHAP) analysis identifies the NCG and total pressure as the dominant variables affecting HTC. These findings align well with experimental physical mechanisms, validating the interpretability of models.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106198"},"PeriodicalIF":3.2,"publicationDate":"2025-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789027","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimentation and simulation research on foreign object wear of cladding-rods interface in high-temperature pressurized water","authors":"Yanke Wen ,&nbsp;Hongyang Chen ,&nbsp;Chuangming Ning ,&nbsp;Yuhan Zheng ,&nbsp;Zhengyang Li ,&nbsp;Zhenbing Cai","doi":"10.1016/j.pnucene.2025.106188","DOIUrl":"10.1016/j.pnucene.2025.106188","url":null,"abstract":"<div><div>The high-speed flow of coolant in a nuclear reactor flushes the fuel rods, causing them to undergo flow-induced vibration (FIV). Additionally, tiny metal impurity particles are captured by the grid as the coolant flows, resulting in fretting wear. These foreign objects are primarily swarf generated during the machining of components in the nuclear reactor. Prolonged fretting can lead to localized wear of the fuel rod cladding, potentially resulting in fuel breakage and the release of fission products, thereby compromising the safety of the nuclear power plant. Foreign object wear is currently an important cause of fuel rod failure, yet little research has been done in this area. The small size of foreign particles renders the experimental studies of foreign object wear challenging. The finite element method (FEM) was employed to simulate the positions and quantities of variously shaped foreign object particles captured in the high-temperature pressurized water environment. In this paper, a new experimental method for foreign object wear was proposed, and the behavior of fretting wear between different foreign objects and cladding tubes is simulated by combined experimental and simulation studies. The study focused on analyzing the effects of particle shape and normal force on abrasion performance. The simulation results indicate that some foreign objects become lodged between the rigid support protrusion and the fuel rod cladding tube, remaining in place despite coolant flushing. Experimental data closely match simulation results at low loads.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106188"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788991","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Multidisciplinary design optimization of a conceptual used nuclear fuel packaging plant","authors":"Grant Minor ,&nbsp;Chantal Medri ,&nbsp;Kamran Behdinan","doi":"10.1016/j.pnucene.2025.106092","DOIUrl":"10.1016/j.pnucene.2025.106092","url":null,"abstract":"<div><div>Multidisciplinary design optimization (MDO) employing a genetic algorithm (GA) is explored as a tool to optimize a conceptual used fuel packaging plant (UFPP), which will be part of a deep geological repository (DGR) for used CANDU reactor fuel. A plant model comprised of seven sequential processing nodes is described. Each node represents a hot cell environment within which steps of the fuel encapsulation process occur. The reliability of the equipment at each processing node in the plant is mathematically modelled considering radiation damage to the limiting sensitive components. The design space includes options for redundant processing cell configurations at each node, which increase capital and operating costs, but improve reliability, reduce downtime, and reduce failure costs. Also considered are the size and capacity of the long-lived used fuel container (UFC) within which the used fuel is encapsulated. The fuel capacity and structural wall thickness of the UFC affect the radiation fields emitted towards the limiting sensitive components at each plant processing node, the lifecycle cost of the equipment assumed at each node, and the UFC material and manufacturing costs. Thus, the disciplinary interdependency between UFC structural design, cost, radiation, and processing node reliability is considered in this novel MDO application for a used nuclear fuel encapsulation facility. The design objective in this study is to minimize lifecycle capital and operating costs for the UFPP and UFC. Considering the input assumptions, an optimum design is found and discussed, and sensitivity studies are performed and evaluated.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106092"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788992","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Uniform approach to validation of codes in application for conservative and realistic accidents modeling","authors":"E.V. Moiseenko,&nbsp;N.A. Mosunova,&nbsp;V.F. Strizhov,&nbsp;D.S. Synitcyn","doi":"10.1016/j.pnucene.2025.106195","DOIUrl":"10.1016/j.pnucene.2025.106195","url":null,"abstract":"<div><div>Modern best-estimate codes can be applied for both conservative and realistic accident modeling. Furthermore, the same experiments can be used for code validation with respect to both these approaches. However, conservative and realistic numerical simulations serve distinct purposes and thus require different validation strategies. For realistic modeling, the goal of validation is to estimate the code's error relative to the “true” value. In contrast, for conservative modeling, validation should demonstrate that the code provides sufficient conservatism or quantify how far the results deviate from conservative values. We propose a method that uses the same set of multivariate simulations, combined with different statistical post-processing, to obtain both types of estimates. These estimates can then be applied to nuclear installations simulations to determine either realistic or conservative values for the parameters of interest.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106195"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788990","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development and validation of a pin-by-pin neutronics/thermal-hydraulics transient coupling code for PWRs based on the SP3 method","authors":"Zhigang Li ,&nbsp;Wenbo Zhao ,&nbsp;Bangyang Xia ,&nbsp;Xiaoming Chai ,&nbsp;Wei Zeng ,&nbsp;Junjie Pan ,&nbsp;Xiafeng Zhou ,&nbsp;Shenglong Qiang ,&nbsp;Wei Lu","doi":"10.1016/j.pnucene.2025.106194","DOIUrl":"10.1016/j.pnucene.2025.106194","url":null,"abstract":"<div><div>Solving the neutron transport equation by using the SP_n method to achieve full-core pin-by-pin calculation represents a significant research direction in nuclear reactor numerical simulation. This work builds upon CORCA-SPn's existing capability to solve the steady-state SP₃ equations using the Two-Node Method combined with Coarse Mesh Finite Difference (CMFD) and Fine Mesh Finite Difference (FMFD). This solution methodology was extended to solve the 3-dimensional(3D) SP₃ space-time kinetics equations in this paper. The backward Euler scheme and theta method are respectively used for the discretization of neutron dynamics SP₃ equations and fuel thermal conductivity equations in CORCA-SPn. Furthermore, leveraging the Multi-physics Object-oriented Reactor Engineering (MORE) framework developed by the Nuclear Power Institute of China (NPIC), the newly developed SP₃ transient transport module is coupled with the existing single-channel thermal-hydraulics module TH1D. This integration enables transient full-core pin-by-pin neutronics/thermal-hydraulics(N-TH) coupling calculation code. The Picard method and the operator splitting(OS) method are respectively employed to govern the iterative procedures for steady-state and transient N-TH coupling calculations in CORCA-SPn. The code's capability for solving SP₃ space-time kinetics problems was validated by using C5G7-TD benchmark. Its transient N-TH coupling capability was validated by using PWR MOX/UO₂ benchmark. The results demonstrate that: 1) CORCA-SPn exhibits good accuracy in solving SP₃ space-time kinetics equations and performing transient N-TH coupled calculations, reliably predicting key transient parameters such as core relative power, reactivity, and fuel Doppler temperature. 2) Validation against the neutron kinetics benchmark of C5G7-TD shows a maximum relative deviation in transient peak power of 6.41 % compared to MPACT. 3) Validation against the control rod ejection benchmark of PWR MOX/UO₂ shows the maximum relative deviation in transient peak power is 20.1 %, the maximum deviation in peak reactivity is −0.05 $, the maximum deviation in fuel Doppler temperature is 3.39 °C, and the maximum deviation in average coolant temperature is 0.19 °C compared to PARCS and Serpent/SCF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106194"},"PeriodicalIF":3.2,"publicationDate":"2025-12-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788989","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}