Progress in Nuclear Energy最新文献

{"title":"Machine learning-enhanced dynamic probabilistic safety assessment for station blackout in Bushehr nuclear power plant","authors":"Mehdi Yarizadeh-Bene ,&nbsp;Mahdi Zangian ,&nbsp;Abdolhamid Minuchehr ,&nbsp;Hamed Kargaran","doi":"10.1016/j.pnucene.2026.106240","DOIUrl":"10.1016/j.pnucene.2026.106240","url":null,"abstract":"<div><div>Dynamic Probabilistic Safety Assessment (D-PSA) faces significant computational challenges when simulating complex accident scenarios with high-fidelity thermal-hydraulic codes, necessitating innovative approaches to balance accuracy and efficiency. This study introduces a machine learning-enhanced Dynamic Event Tree (DET) module that improves risk analysis quantification by integrating deep learning and Support Vector Machine (SVM). This module utilizes a FORTRAN-based submodule for automated generation and parallel execution of thermal hydraulic model input files, enabling concurrent simulation of several of dynamic scenarios while a random sampling strategy ensures comprehensive coverage of failure sequences with minimized training data requirements. In this research a Station blackout (SBO), as initiating event at Bushehr nuclear power plant (BNPP) is considered for thermal-hydraulic model benchmark. Consequently, dynamic effect of diesel generators recovery in different time steps are evaluated. As a result, SBO turns into the loss of offsite power (LOOP) accident. The proposed module investigates LOOP accident for comprehensive risk analysis at BNPP. The results show that the machine learning architecture developed achieves good predictive performance, surpassing at least 97 % accuracy in classifying core damage states and reducing scenario evaluation time. High-resolution dynamic modeling combined with computational feasibility in this module represents fast and effective method in nuclear safety analysis.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106240"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977698","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":"Numerical analysis on heat transfer and flow characteristics of supercritical CO2 laminar mixed convection in horizontal non-uniform heat flux tubes","authors":"Shengqian Wei ,&nbsp;Yao Fu ,&nbsp;Yang Yang ,&nbsp;Xiangzhou Cai","doi":"10.1016/j.pnucene.2025.106217","DOIUrl":"10.1016/j.pnucene.2025.106217","url":null,"abstract":"<div><div>Molten salt fast reactor (MSFR) cooled directly by supercritical CO₂ has a compact design, high power outputs, and promising development opportunities. This research will offer essential insights for thermal design and optimization of MSFR. The thermos-hydraulic properties of supercritical CO₂ undergoing laminar mixed convection are numerically simulated in this study. The supercritical CO₂ is heated by non-uniform heat flux similar to the axial power distribution in MSFR. The impact of buoyancy force, heat flux, and inlet Reynolds number on the thermos-hydraulic characteristics of supercritical CO₂ was assessed. For a non-uniform heat flux, the Nusselt number (<em>Nu</em>) of mixed convections exceeds that of forced convection by up to 167.62 % while the Fanning friction coefficient increases by 75.99 %. As the heat flux and inlet Reynolds number increase, the local <em>Nu</em> of the mixed convection increases, albeit at a decreasing rate. With the increase of heat flux, the heat transfer coefficient increases by 53.64 % for mixed convection with a non-uniform heat flux. The Fanning friction coefficient increases with higher heat flux and decreases with higher inlet Reynolds number. With the increase of the inlet Reynolds number, the Fanning friction coefficient reduces to 56.51 % for uniform heat flux. Uneven heat flux causes a higher local temperature peak, alters the curve of the local <em>Nu</em>, extends the inlet section, and affects significantly the Fanning friction factor at the inlet (z/D &lt; 12).</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106217"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884776","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":"Research on the influence mechanism of non-uniform inflow on the pressure pulsation of the Reactor Coolant Pump","authors":"Yun Long,&nbsp;Zhenguo Wu,&nbsp;Mingyu Zhang,&nbsp;Xi'an Guo,&nbsp;Yuan Xu,&nbsp;Chenbiao Tian,&nbsp;Yingying Zheng,&nbsp;Xiangrong Luo","doi":"10.1016/j.pnucene.2025.106208","DOIUrl":"10.1016/j.pnucene.2025.106208","url":null,"abstract":"<div><div>Based on a series of previous research findings from our team, the special direct-connection structure between the Steam Generator and the Reactor Coolant Pump causes asymmetry in the velocity distribution of the inlet inflow, and this flow pattern is defined as non-uniform inflow. This asymmetric velocity distribution characteristic at the inlet will further lead to inlet velocity distortion at the pipe cross-section. To reveal the evolution law of pressure pulsation and clarify the action mechanism of distorted inflow, this study carries out unsteady numerical simulations based on the SP/SGSE-RCP model. By innovatively analyzing the coupling relationship between components, it elucidates the internal mechanism underlying changes in pressure pulsation characteristics, providing a new perspective for investigating complex flows in RCPs. Multiple monitoring points were placed in the coupling model, and Fast Fourier Transform was applied to analyze the simulated pressure pulsation characteristics. Results show that non-uniform inflow significantly enhances RCP pressure pulsation intensity and amplifies the impeller radial force. Furthermore, it shifts the dominant frequency in impeller channels from 246.67 Hz to 24.67 Hz, while the dominant frequencies of the diffuser and casing remain consistent with those under uniform inflow. Additionally, non-uniform inflow increases the harmonic components of pressure pulsation in the casing, especially in the “volute tongue” area. This study investigates the mechanism by which non-uniform inflow regulates RCP pressure pulsation, offering critical insights for future RCP optimization.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106208"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884611","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":"Data-driven prediction of HTR-PM bottom reflector maximum temperature using recurrent neural networks","authors":"Ziwen Zhang ,&nbsp;Zhongkai Sheng ,&nbsp;Bing Xia ,&nbsp;Ding She ,&nbsp;Zuoyi Zhang","doi":"10.1016/j.pnucene.2025.106200","DOIUrl":"10.1016/j.pnucene.2025.106200","url":null,"abstract":"<div><div>In pebble-bed high-temperature gas-cooled reactors (HTGRs), accurate monitoring and prediction of the maximum bottom reflector temperature are critical for safety margins and operations. To complement traditional physics-based simulation methods, we propose a data-driven time-series approach based on recurrent neural networks (RNNs) to predict the maximum temperature at the bottom reflector funnel location directly from operation histories. Using eight fixed operational variables as inputs and the measured maximum bottom reflector temperature as output, with additional detector-based features in optimal configurations, the RNN achieves root-mean-square errors (RMSE) of 13-21 °C when compared to measured temperature values. We report configurations with stationarized power steps and with additional detector signals, and demonstrate cross-module generalization capability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106200"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884778","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":"Effective obsolescence management for improving safety of existing nuclear power plants","authors":"Kenta Murakami ,&nbsp;Naoto Sekimura","doi":"10.1016/j.pnucene.2025.106199","DOIUrl":"10.1016/j.pnucene.2025.106199","url":null,"abstract":"<div><div>This paper reviews diverse practices on safety improvements of existing nuclear fleets from the viewpoint of obsolescence management. First, the characteristics of obsolescence management in nuclear power were categorized and were compared with those in other industries. Next, for each category, the corresponding regulatory tools in multiple states were compared. Then, representative practices in the six technical areas, some of which are not traditionally recognized as obsolescence management, are presented and their characteristics are summarized. The parties involved in the obsolescence management extend beyond the typical “licensee versus regulator” structure, so that a mechanism to pay incentives to stakeholders for their efforts to improved safety is needed. The relationship between the obsolescence and quality management system is discussed, and the importance of “marketing of safety”, as a countermeasure of knowledge obsolescence, is discussed.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106199"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884775","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":"NURBS-enhanced control volume finite element spatial discretisation methods for the steady-state multigroup neutron diffusion equation","authors":"J. Trainor ,&nbsp;M.D. Eaton ,&nbsp;J. Kópházi ,&nbsp;S.G. Wilson ,&nbsp;C. Latimer ,&nbsp;L. Smith ,&nbsp;D. Baker ,&nbsp;I. Jordan","doi":"10.1016/j.pnucene.2025.106197","DOIUrl":"10.1016/j.pnucene.2025.106197","url":null,"abstract":"<div><div>The paper describes a novel spatial discretisation termed the NURBS-enhanced control volume finite element Method (NECVFEM) and applies it to the solution of the steady-state multigroup neutron diffusion equation. The NECVFEM combines the geometric flexibility of FEM with the local, and global, conservation properties of finite volumes and the exact geometric representation associated with Non-Uniform Rational B-Splines (NURBS). In this paper, the NECVFEM and CVFEM discretisations are used to solve three nuclear reactor physics benchmark verification test cases. The numerical accuracy, and computational efficiency, of the NECVFEM and CVFEM have been compared to standard Lagrangian FEM. The Lagrangian FEM has utilised both volume and surface preserving approaches for curvilinear computational domains. The NECVFEM method has been shown to be highly effective in reducing the geometric errors associated with modelling curvilinear computational domains. It has achieved lower numerical errors than the CVFEM across all spatial refinements and benchmark verification test cases. In addition, the NURBS-enhanced methods have been shown to produce geometric errors that are lower, or of similar magnitude, to that of the volume and surface preserving methods while avoiding the need for complex mesh modification steps. A comparison of computational solution times associated with each discretisation method showed that the NECVFEM was more computationally efficient than the CVFEM and comparable to the FEM, particularly for linear spatial discretisations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106197"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977695","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":"Innovations and challenges in natural circulation for advanced nuclear reactors: A comprehensive review of passive safety and future research opportunities","authors":"Samah A. Albdour ,&nbsp;Imran Afgan ,&nbsp;Yacine Addad","doi":"10.1016/j.pnucene.2025.106193","DOIUrl":"10.1016/j.pnucene.2025.106193","url":null,"abstract":"<div><div>Natural circulation systems are fundamental to passive safety in nuclear power plants, enabling effective heat removal without active mechanical intervention. This review explores the thermal-hydraulic behavior of natural circulation phenomena across diverse reactor types, including land-based water-cooled reactors, liquid metal-cooled reactors, and floating nuclear power plants (FNPPs). Drawing on extensive experimental and computational studies, the review highlights critical factors influencing system performance, including flow stability, heat transfer efficiency, and scaling effects. The discussion emphasizes state-of-the-art Computational Fluid Dynamics (CFD) and system codes, for modeling natural circulation and addressing challenges like asymmetric flows, thermal stratification in liquid metal coolants, and dynamic instabilities in FNPP environments. This review highlights future research directions, emphasizing the development of hybrid systems combining active control mechanisms (e.g., machine learning-based sensors and feedback loops) with passive safety features. Key areas include marine simulation facilities for FNPPs, advanced fluid-structure interaction models, and the application of innovative materials like graphene and nanofluids. By synthesizing advancements in experimental techniques, computational methods, and design optimization, while proposing actionable research directions, this work provides a roadmap for the next generation of natural circulation systems. These findings support the safe and efficient operation of nuclear reactors in diverse environments and contribute to global energy sustainability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106193"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789076","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":"Uncertainty aware unsupervised fault diagnosis of PWR nuclear power plant using KNN and SHAP method","authors":"Fazle Haseeb ,&nbsp;Minjun Peng ,&nbsp;Furqan Arshad ,&nbsp;Wasiq Ali","doi":"10.1016/j.pnucene.2025.106191","DOIUrl":"10.1016/j.pnucene.2025.106191","url":null,"abstract":"<div><div>Machine learning based fault diagnosis methods in nuclear power plants (NPPs) often exhibit lack of labeled fault data in sufficient quantity, thus posing a significant challenge for the supervised fault diagnosis. This paper presents a novel, uncertainty-aware, and unsupervised fault diagnosis framework for pressurized water reactor (PWR) type NPP. For this purpose, it integrates the k-nearest neighbors (KNN) method with shapley additive explanations (SHAP) analysis, not only to enhance the model interpretability but also to ensure its reliability. The KNN model, trained on steady-state operational data, firstly detects anomalies, while SHAP analysis subsequently performs the fault classification in an unsupervised manner. For the purpose of fault classification, feature contributions are computed for each detected anomaly, thus providing an interpretable features pattern. This pattern information is further analyzed and transformed into an entropy-based metric, which is calculated from the normalized SHAP values. This metric enables a transparent classification between the sensor and process faults. Regarding the sensor faults, this methodology further provides the identification of the faulty sensor through the analysis of the highest contributing feature. In this way, explainable artificial intelligence (XAI) has been utilized not only for model interpretation but has also played its role in the fault diagnosis process. Additionally, a bootstrap resampling approach has been incorporated for the estimation of model uncertainty, thus providing the confidence measures that support safer and risk-informed decisions. Performance of the proposed methodology has been evaluated against various sensor and process faults, not only through clean data but also against noisy data. An inter-comparison among various anomaly detection methodologies within the proposed framework has also been presented. The testing results demonstrate that the proposed method achieves 99.56% ± 0.051% (mean ± standard deviation) classification accuracy across 100 bootstrap iterations, and also shows resilience under noisy conditions. This fully unsupervised, interpretable, and uncertainty-informed framework thus offers a significant advancement for reliable fault diagnosis in safety-critical nuclear power plant operations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106191"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789077","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 the dynamic characteristics of fluid-structure interaction of immersed double-layer thin shells in liquid metal fast reactors","authors":"Yubo Zhang ,&nbsp;Donghao Li ,&nbsp;Chaofan Zhang ,&nbsp;Hewei Wang ,&nbsp;Yuqi Zhou ,&nbsp;Daogang Lu ,&nbsp;Yu Liu","doi":"10.1016/j.pnucene.2025.106218","DOIUrl":"10.1016/j.pnucene.2025.106218","url":null,"abstract":"<div><div>The primary vessel of a pool-type sodium-cooled fast reactor contains a large amount of liquid sodium and various complex internal structures, such as heat exchangers and main pump support cylinders. The main pump support cylinder and its thermal shield are multi-layer thin-walled structures that can be regarded as a partially submerged, coaxial double-layer flexible shell. Under seismic conditions, significant fluid–structure interaction (FSI) occurs between the liquid sodium and the main pump support assembly, which alters the system's vibration characteristics and affects the structural safety of the reactor. In this study, a simplified and scaled-down test model of the main pump support cylinder and thermal shield was established. Tests were conducted using both an exciter and a shaking table under different fluid gap conditions to obtain the added mass and added damping. The applicability of the two-dimensional FSI theoretical method for partially submerged structures was then discussed, and it was found that this method exhibits limited accuracy for out-of-phase modes and low-order in-phase modes. Finally, modal simulations were conducted using the commercial ANSYS software to verify the accuracy of the finite element method (FEM) in predicting the added mass effects associated with fluid–structure interaction in such structures.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106218"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884610","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 of (alpha, n) neutron source and spectrum calculation capabilities in the NUIT code","authors":"Bo Tan,&nbsp;Yutong Wen,&nbsp;Ding She,&nbsp;Lei Shi","doi":"10.1016/j.pnucene.2025.106213","DOIUrl":"10.1016/j.pnucene.2025.106213","url":null,"abstract":"<div><div>This paper presents the development of capabilities for calculating (alpha, n) neutron sources and spectra within the NUIT (Nuclide Inventory Tool) code, including four computational models: the homogeneous material model, the beam model, the interface model, and the multi-region model. Crucially, the novel multi-region model proposed in this work addresses the limitations of the three-region model in the Los Alamos National Laboratory’s SOURCES code, thereby allowing for the accurate treatment of more complex geometric configurations. The results show that the NUIT code agrees well with the SOURCES code for the homogeneous material, beam, and interface models, with discrepancies below 2.0 %. Furthermore, the validity and accuracy of the proposed multi-region model are also verified by appropriate numerical cases.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106213"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884613","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}