Progress in Nuclear Energy最新文献

{"title":"Assessment and experimental validation of the new wall boiling heat transfer model under different boundary conditions","authors":"Xiang Zhang ,&nbsp;Jie Wan ,&nbsp;Nikolai Rensch ,&nbsp;Xu Cheng","doi":"10.1016/j.pnucene.2025.106043","DOIUrl":"10.1016/j.pnucene.2025.106043","url":null,"abstract":"<div><div>Flow boiling widely exists in industrial systems because of its high heat transfer capability. In the previous study, we proposed a new heat transfer model for wall boiling through mechanism analysis (Zhang et al., 2024). As a newly developed model, it is quite important to explore the impact of different boundary conditions on model performance to provide potential directions for further improvement. In this paper, a series of flow boiling experiments for R134a within a wide range of boundary conditions (<em>q</em>_<em>w</em> = 80–260 kW/m², <em>G</em> = 1000–2000 kg/(m²·s), <em>P</em> = 11–17 bar, <span><math><mrow><mo>Δ</mo></mrow></math></span> <em>T</em>_in,sub = 19–38 K) were carried out in a 10 mm-diameter vertical tube, providing more than 900 data points. The developed wall boiling model was validated using these experimental data for R134a boiling, as well as referenced water boiling and n-Perfluorohexane boiling experimental data in both pipes and rectangle channels. According to the results, the mean relative error of the wall superheat between our experimental data and the model predictions is ±31.5 %. In addition, an analysis was conducted to examine the influence of different boundary conditions on the accuracy of the developed model. The reasons for the discrepancy of the calculated results were analyzed and further improvements should be considered in the model.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106043"},"PeriodicalIF":3.2,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047210","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 research on thermal-hydraulic characteristics of generation Ⅲ PWR under LOCA conditions","authors":"Liangguo Li ,&nbsp;Qianhua Su ,&nbsp;Zongbao Nan ,&nbsp;Jianming Yu ,&nbsp;Donghua Lu ,&nbsp;Xianghui Lu ,&nbsp;Yingwei Wu ,&nbsp;Suizheng Qiu ,&nbsp;Xin Wang ,&nbsp;Jiyong Liu","doi":"10.1016/j.pnucene.2025.106034","DOIUrl":"10.1016/j.pnucene.2025.106034","url":null,"abstract":"<div><div>China has developed the HPR1000, an advanced Generation III pressurized water reactor (PWR) that integrates active and passive safety systems to achieve enhanced nuclear power plant safety. Among all design basis accidents (DBAs), the loss of coolant accident (LOCA) represents the most severe challenge for the reactor safety analysis. To investigate LOCA characteristics, the HPR1000 integral test facility (HITE) was designed using the hierarchical two-tiered scaling (H2TS) methodology, accurately simulating the three-loop configuration of the reference reactor. The facility was employed to conduct a LOCA experiment with 4 % cold leg area. Then, the detailed analysis of the resultant system transients and key thermal-hydraulic phenomena were obtained based on the experimental and simulation results. Primary system pressure was effectively reduced during medium pressure rapid cooling (MPRC) operation through steam discharge via VDA under LOCA conditions. The MPRC system achieved a rapid cooldown rate of 500 °C/h and depressurization, enabling timely RIS activation. Both core fluid temperatures and fuel rod cladding temperatures decreased during the accident. The collapsed water level of RS is recovered until reaches the bottom of CL. The experimental results show that the system configuration and capacity is rational in coping with LOCA accident. The thermal stratification phenomenon is observed and discussed in the pipes of primary system. It is evident that the overall variation trend of the pressure obtained from the calculation is generally consistent with the experimental measurements. The findings are expected to improve the understanding of LOCA in HPR1000.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106034"},"PeriodicalIF":3.2,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047171","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 novel reduced-order model and two-stage optimization strategy for high-efficiency predicting thermal-hydraulic behavior in a 6×6 fuel bundle","authors":"Guangyun Min ,&nbsp;Xiuzhong Shen ,&nbsp;Laishun Wang ,&nbsp;Naibin Jiang","doi":"10.1016/j.pnucene.2025.106032","DOIUrl":"10.1016/j.pnucene.2025.106032","url":null,"abstract":"<div><div>Achieving fast and low-cost computation of the thermal-hydraulic flow field inside a reactor is of great significance for understanding the reactor's thermal-hydraulic characteristics and ensuring its safety. Based on the Fluent scripting technique, a total of 216 cases with varying inlet velocities, inlet temperatures and power densities were simulated, forming the dataset for model training. A reduced-order model (ROM), referred to as proper orthogonal decomposition (POD)-surrogate model (SM)-ROM, was constructed by combining the POD method with SM. The model takes inlet velocity, inlet temperature, and power density as inputs and outputs the corresponding POD mode coefficients. The proposed POD-SM-ROM can enable rapid prediction of flow fields and temperature fields under non-sample points operating conditions by reconstructing the flow fields and temperature fields from the predicted POD mode coefficients and POD mode. In addition, a novel two-stage optimization strategy for the POD-SM-ROM was proposed to further reduce computational time while preserving accuracy. The ROM and two-stage optimization framework developed in this study enables efficient and rapid prediction of thermal-hydraulic behavior in nuclear reactors. This advancement not only enhances computational efficiency but also supports timely and reliable reactor safety assessments. All the codes in this paper can be found at our GitHub link: (<span><span>https://github.com/guangyunmin/A-novel-data-driven-reduced-order-model</span><svg><path></path></svg></span>).</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106032"},"PeriodicalIF":3.2,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047207","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 and numerical analysis of sloshing in annular sectored containers under tangential direction excitation","authors":"Daogang Lu ,&nbsp;Fei Xie ,&nbsp;Feifan Zhang ,&nbsp;Yu Liu ,&nbsp;Yuxuan Zhu ,&nbsp;Yixian Zhou","doi":"10.1016/j.pnucene.2025.106046","DOIUrl":"10.1016/j.pnucene.2025.106046","url":null,"abstract":"<div><div>In many third-generation pressurized water reactor (PWR) nuclear power plants, large annular passive cooling water tanks are typically installed atop the containment structure to assist in dissipating residual heat from the reactor during station blackout and other accident conditions. These tanks are partitioned by radial walls into multiple sectored compartments. Due to their elevated position, these annular sectored tanks exhibit strong seismic responses. In particular, under long-period ground motions, liquid sloshing inside the tanks can impose additional hydrodynamic loads on the structure, potentially compromising its structural integrity. Therefore, a thorough investigation into the sloshing behavior of such tanks under seismic excitation is essential for ensuring their seismic safety. At present, experimental studies on annular sectored tanks under large-amplitude sloshing conditions are very limited. Moreover, the applicability and accuracy of existing numerical methods in handling such complex scenarios require further validation and assessment. To address this, the present study conducts both experimental and numerical investigations of annular sectored tanks. A scaled shaking table test is performed to systematically analyze the modal characteristics and transient response of sloshing under tangential excitation. The effects of excitation intensity, tank geometry, and excitation direction on key parameters such as wave height and hydrodynamic pressure are evaluated. In terms of numerical simulation, three approaches are employed: a finite element-based acoustic-structure coupling method, a CFD-based Volume of Fluid (VOF) method, and a two-way fluid-structure interaction (FSI) method. The simulation results are compared with experimental data to verify the accuracy and applicability of each approach under various conditions. The findings of this study provide valuable insights for seismic response assessment and structural optimization of annular sectored passive cooling tanks.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106046"},"PeriodicalIF":3.2,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047170","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":"Transient numerical studies on Departure from Nucleate Boiling (DNB) crisis in petal-shape fuel rod bundle assembly","authors":"Lipeng Du ,&nbsp;Xiang Chen ,&nbsp;Qi Cheng ,&nbsp;Wenchao Zhang ,&nbsp;Jianchuang Sun ,&nbsp;Weihua Cai","doi":"10.1016/j.pnucene.2025.106025","DOIUrl":"10.1016/j.pnucene.2025.106025","url":null,"abstract":"<div><div>The DNB crisis may cause fuel clad damage or melting which is a reactor safety concern for radiation release. For subcooled boiling, the decisive factor of the DNB boiling crisis is the heat flux. Therefore, it is particularly important to predict critical heat flux. Apply the models of Eulerian two-phase flow and the RPI wall boiling, the DNB in the petal-shaped fuel assembly is simulated. The Critical Heat Flux (CHF), the position of boiling crisis, and temperature field characteristics and void fraction distribution near position of boiling crisis are studied under the transient conditions, such as the sudden drop of the flow rate and the sudden rise of the thermal power. The results show that an increase in the power change ratio induces a proportional rise in CHF, when fuel rod power sudden increases. Wall temperature is higher, when the inlet velocity drops suddenly, and wall temperature increases with inlet velocity drop more quickly, when heating time is the same. The position of the DNB appears at FUEL3, with the two sets of cases; in the upstream and downstream of DNB, wall heat flux undergoes redistribution, peaking at the DNB site before subsequent reduction.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106025"},"PeriodicalIF":3.2,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047208","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":"Optimizing deep neural networks for nuclear power plant temperature estimation: A study on feature importance and outlier detection","authors":"Bernardo M. Caixeta ,&nbsp;João V.S. A. Guimaraes ,&nbsp;Marcelo C. Santos ,&nbsp;Matheus C. Silva ,&nbsp;Andressa S. Nicolau ,&nbsp;Roberto Schirru ,&nbsp;Da Silva M. Candeias ,&nbsp;Muzitano G. Frazão ,&nbsp;Justino M. Castro","doi":"10.1016/j.pnucene.2025.106039","DOIUrl":"10.1016/j.pnucene.2025.106039","url":null,"abstract":"<div><div>Extending the operational life of the Angra 1 Nuclear Power Plant (NPP) requires an accurate estimation of historical temperature exposure for equipment within the containment area to assess the aging and degradation of critical components. This assessment is essential for extending the plant's license by 20 years. Since Mobile Temperature Sensors (MTSs) were installed only in 2015, this study employs Deep Neural Networks (DNNs), including Deep Rectifier Neural Networks (DRNNs), Convolutional Neural Networks (CNNs), and Long Short-Term Memory Networks (LSTMs), to infer historical temperature data before MTS deployment. The DNNs utilize time series data from Plant Fixed Sensors (PFS), monitored by the Angra 1 Integrated Computer System (SICA), as inputs. Feature importance and outlier detection methods are investigated to enhance DNN performance. Feature importance techniques, such as XGBoost, Random Forest, Principal Component Analysis (PCA), and outlier detection methods, including autoencoders, DBSCAN, and isolation forest, are evaluated. Results indicate that preprocessing significantly improves model accuracy. For instance, PCA without outlier detection combined with a CNN achieved a Mean Absolute Error (MAE) of 3.194, whereas the integration of Random Forest and XGBoost for feature importance with DBSCAN for outlier detection and a CNN reduced the MAE to 0.497.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106039"},"PeriodicalIF":3.2,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047197","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 and numerical study on thermal-hydraulic characteristics of water-cooled fuel assembly irradiation device for HFETR","authors":"Liangqian Fu ,&nbsp;Xufeng He ,&nbsp;Shaojie Tan ,&nbsp;Songbai Cheng ,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.106035","DOIUrl":"10.1016/j.pnucene.2025.106035","url":null,"abstract":"<div><div>In this paper, in order to study the thermal-hydraulic behavior of fuel assembly irradiation device in HFETR (High-Flux Engineering Test Reactor), a fuel assembly irradiation device simulator with narrow channels and slit was developed. The thermal-hydraulic characteristics of the fuel assembly irradiation device was experimentally and numerically investigated. In the experiment, under the benchmarking high and low flow conditions, the overall heat leakage rate of the device is 1.54 % and 0.38 %, respectively, indicating that the thermal insulation performance of the experimental device is good. Under the two flow conditions, the corresponding total pressure drop in the experimental body is about 38 kPa and 11 kPa, respectively. Compared with the experiment, the maximum experimental body pressure drop error of the calculated results is 9.58 %. The calculated results show that there is a drastic exchange of velocities between the fluids during the movement of the fluid in the downcomer section. The flow velocity inside the slit channel is larger than that in the slit inlet and outlet area, and the velocity gradient at the slit outlet is larger. The existence of the slit makes a part of the fluid enter the outflow channel of the electric heating section, accounting for 2.4 %. The fluid mixing effect of the square box in the device is obvious, which is conducive to the flow distribution of the fluid into the heating channel. These findings provide critical insights for the design optimization of nuclear fuel irradiation devices in HFETR, particularly offering direct guidance for temperature measurement point arrangement, enhancing the thermal-hydraulic parameter measurement accuracy, and optimizing the narrow-channel flow control.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106035"},"PeriodicalIF":3.2,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027302","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 investigation of high-temperature heat pipe bundle failure scenario for heat pipe cooled reactor","authors":"Jinyuan Wang,&nbsp;Panxiao Li,&nbsp;Chenglong Wang,&nbsp;Wenxi Tian,&nbsp;Guanghui Su,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.106033","DOIUrl":"10.1016/j.pnucene.2025.106033","url":null,"abstract":"<div><div>The heat pipe reactor is an innovative nuclear reactor technology that employs alkali-metal heat pipes to transfer heat efficiently. However, during operation, heat pipes face significant thermo-mechanical loads, which can lead to failures and reduced cooling performance. While numerous studies have conducted numerical simulations to investigate potential failure scenarios and assess the safety of heat pipe reactors, there remains a lack of validation against experimental data. This study evaluates the feasibility of using the thermal resistance network method for simulating accident scenarios based on experimental data from high-temperature heat pipe bundle failure tests. Results show that individual performance differences among heat pipes must be included in simulations; neglecting this can underestimate the maximum matrix temperature. Additionally, modeling failed heat pipes should not assume adiabatic boundary conditions but rather increase thermal resistance appropriately to preserve partial heat transfer capability. Using adiabatic conditions results in a higher maximum matrix temperature compared to applying increased thermal resistance. When the performance differences among individual heat pipes and the corresponding thermal resistance variations of failed heat pipes are taken into account, this approach yields a more accurate representation of the temperature distribution.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106033"},"PeriodicalIF":3.2,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145047196","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":"Artificial neural networks for nuclear power: Analysis of boiling water reactor control rod patterns in the startup range","authors":"Alexey Cherezov ,&nbsp;Victor Fournier ,&nbsp;Alexander Vasiliev ,&nbsp;Jiri Dus ,&nbsp;Hakim Ferroukhi","doi":"10.1016/j.pnucene.2025.105997","DOIUrl":"10.1016/j.pnucene.2025.105997","url":null,"abstract":"<div><div>The start-up phase of boiling water reactors poses a safety challenge due to the risk of control rod malfunction. During this phase, control rods are withdrawn sequentially to achieve reactor criticality, and then the nominal power. However, a failed control rod may unexpectedly drop, causing a surge of positive reactivity that could trigger a prompt criticality accident, presenting serious safety risks. The reactivity worth of a dropped rod depends on the positions of other rods, making it crucial to avoid dangerous configurations. Evaluating all possible rod arrangements (<span><math><mrow><mo>∼</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>10</mn></mrow></msup></mrow></math></span> and more), is infeasible, and only a small fraction of these configurations (<span><math><mrow><mo>∼</mo><mn>0</mn><mo>.</mo><mn>001</mn><mtext>%</mtext></mrow></math></span>) represents safety concerns. To address this, we propose a novel approach using artificial neural networks to predict the reactivity of control rod configurations. This method enables the rapid and efficient identification of limiting patterns, providing a practical solution to enhance reactor safety during start-up operations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 105997"},"PeriodicalIF":3.2,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020059","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":"Design of rectangular single-phase natural circulation loop based on working fluid and geometry parameters to enhancement stability and mass flow rates: A research review","authors":"Roy Waluyo ,&nbsp;Deendarlianto ,&nbsp;Indarto ,&nbsp;Dwi Yuliaji ,&nbsp;Veronica Indriati Sri Wardhani ,&nbsp;Akhmad Afandi ,&nbsp;Muhammad Ganjar Putra ,&nbsp;Ryan Oktaviandi ,&nbsp;Sunandi Kharisma ,&nbsp;Shendy Akbar Maryadi ,&nbsp;Achilleus Hermawan Astyanto ,&nbsp;Mulya Juarsa","doi":"10.1016/j.pnucene.2025.106037","DOIUrl":"10.1016/j.pnucene.2025.106037","url":null,"abstract":"<div><div>This literature review examines diverse experimental and numerical investigations of single-phase natural circulation loops, with particular emphasis on rectangular configurations. The analysis focused on two critical aspects: mass flow rate and instability phenomena. The examination of instability is categorized into three main sections: (1) The influence of geometric parameters and operational variables on single-phase natural circulation loop stability, (2) Methods for improving single-phase natural circulation loop stability through loop configuration modifications and component additions, and (3) stability mapping approaches for single-phase natural circulation loops. The study also explores how nanofluid implementation as a working medium affects mass flow characteristics and system stability. Our paper synthesizes contemporary developments in natural circulation loop studies, drawing parallels with passive cooling mechanisms in nuclear facilities. We seek to comprehensively assess present conditions and obstacles in enhancing mass flow and stability characteristics within single-phase natural circulation loops. Additionally, this analysis offers guidance for developing rectangular loops with optimized configurations, operational protocols, and geometric specifications.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106037"},"PeriodicalIF":3.2,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020060","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}