Progress in Nuclear Energy最新文献

{"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":"2026-03-01","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":"Experimental investigation of turbulent flow in a rod bundle with flow blockage","authors":"Denggao Chen,&nbsp;Jingliang Bi,&nbsp;Yanping Huang,&nbsp;Dewen Yuan,&nbsp;Yuanfeng Zan,&nbsp;Jianjun Xu","doi":"10.1016/j.pnucene.2025.106215","DOIUrl":"10.1016/j.pnucene.2025.106215","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106215"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884668","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 an improved turbulent heat flux model for the heat transfer of supercritical fluids","authors":"Abdullah Alasif ,&nbsp;Andrea Pucciarelli ,&nbsp;Afaque Shams","doi":"10.1016/j.pnucene.2025.106190","DOIUrl":"10.1016/j.pnucene.2025.106190","url":null,"abstract":"<div><div>For decades, enhancing coolants or working fluid properties has been a main research interest. Operating fluids above their critical point enhance their thermophysical properties, making them suitable for advanced thermal systems. These fluids are referred to as supercritical fluids (SCF). SCF can be utilized in many industrial applications, such as jet propulsion systems, conventional power cycles, and nuclear reactors. However, the inclusion of SCF in such applications is restricted by the challenging heat transfer behavior driven by the drastic change in thermophysical properties. Predicting the heat transfer in such fluids is essential for the safe operation of thermo-sensitive systems like nuclear reactors. Many attempts have been made in open literature to generate numerical methods for predicting the heat transfer of SCF; however, the proposed methods lacked generality and reasonable accuracy. The current work presents an improved advanced turbulent heat flux model in the framework of the Reynolds-Averaged Navier-Stokes (RANS) approach for predicting the heat transfer of SCF. In this regard, an Algebraic Heat Flux Model (AHFM) has been considered as a base for the model. Although the AHFM modeling approach is advanced, it has never reached the level of being used in the generalized form to cater to a wide range of industrial applications. Nevertheless, in this research work, a wide range of numerical studies have been performed to study the complex heat transfer phenomena in SCF. In total, 45 different upward flow cases were considered under various conditions. Existing standard and advanced turbulent heat flux models were found to inadequately predict heat transfer. Therefore, a new formulation for AHFM has been proposed, hereafter referred to as AHFM-APS, to accurately predict the heat transfer in SCF at different conditions. This new model has shown promising results and has significantly improved the accuracy of heat transfer prediction to a wide range of flow conditions.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106190"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145739067","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":"CFD analysis of heat transfer in homogeneous and stratified corium pools for In-Vessel Retention assessment","authors":"Luteng Zhang ,&nbsp;Zhiguo Xi ,&nbsp;Zaiyong Ma ,&nbsp;Wan Sun ,&nbsp;Qiang Lian ,&nbsp;Simiao Tang ,&nbsp;Longxiang Zhu ,&nbsp;Liangming Pan ,&nbsp;Meiyue Yan","doi":"10.1016/j.pnucene.2025.106192","DOIUrl":"10.1016/j.pnucene.2025.106192","url":null,"abstract":"<div><div>This study conducts the numerical investigation into the heat transfer characteristics of corium pools within the reactor pressure vessel lower head under severe accident conditions. Utilizing the open-source CFD platform OpenFOAM, simulations integrate the large eddy simulation approach for turbulence and the enthalpy-porosity model for solidification phase change. The model is validated against benchmark experiments, including Gau's melting case and COPRA experiment. The research systematically analyzes and compares the thermal-hydraulic behavior of single-layer, two-layer, and three-layer corium pool configurations under different top boundary conditions (radiation vs. cooling). Results indicate that pool stratification significantly alters internal temperature distributions and vessel wall heat fluxes. The presence of metallic layers induces pronounced thermal focusing effect, leading to localized heat flux peaks that challenge the success of In-Vessel Retention strategy, while top cooling is shown capable to effectively mitigate this risk. This work provides critical insights for the safety assessment of severe accident in nuclear reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106192"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145739065","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":"Fuel performance Sobol sensitivity analysis for a pressurized water reactor with high burnup fuel during flexible operation using BISON","authors":"Isabelle O. Lindsay,&nbsp;Nicholas R. Brown","doi":"10.1016/j.pnucene.2025.106183","DOIUrl":"10.1016/j.pnucene.2025.106183","url":null,"abstract":"<div><div>Many currently operating nuclear power plants in the United States are at risk of early closure, regardless of their safety and availability records. To preserve this source of clean baseload energy, nuclear power can be used in Integrated Energy Systems or in flexible operation, which can present fuel performance challenges. There is a need to understand flexible operation parameters' impact on fuel safety related metrics, based on the United States Nuclear Regulatory Commission's (NRC) guidance. The fuel safety related metrics addressed are fuel temperature, cladding temperature, plenum pressure, cladding hoop strain, and fission gas release. This work presents a fuel performance sensitivity study, applying cyclic loading for a &gt;5 % U-235 enrichment, annular fuel design with homogenously blended gadolinium. This work uses a Sobol sampling method to develop two and three cycles' worth of power histories with an average discharge burnup of 64.4 GWd/MTU, a life-time average linear heat rate (LHR) of 14.39 kW/m, a range of 0.5 %–5 % rated power ramp rates, a range of 30 %–80 % power scaling factors, and a range of 8-h – 8-day power intervals. In all cases, ramp rate and interval length were the least influential input parameters, where life-average LHR and power scaling factors were highly influential input parameters. The output parameters of interest were also observed in regard to the maximum LHR experienced by the fuel pin. In both fuel cycle analyses, fuel pins began to exceed the pressure limit at approximately 20 kW/m LHR and then began to exceed cladding hoop strain limits at approximately 22 kW/m LHR. If seeking to limit fuel performance metrics to current NRC guidance, pin plenum pressure to below 15.5 MPa and cladding hoop strain to below 1 %, limiting the maximum LHR could be effective in addition to influential parameters from the Sobol analysis. This analysis offers a novel understanding of key safety-related metrics for fuel performance in flexible power operation. An area of importance for future investigation is an in-depth pellet-cladding interaction (PCI) analysis.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106183"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788972","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":"2026-03-01","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 of a risk profile for accident sequences considering release-to-impact correlation","authors":"Yunho Kim,&nbsp;Jaehyun Cho","doi":"10.1016/j.pnucene.2026.106242","DOIUrl":"10.1016/j.pnucene.2026.106242","url":null,"abstract":"<div><div>Following the Three Mile Island accident, risk has become a fundamental concept in evaluating nuclear power plant (NPP) safety. Probabilistic Safety Assessment (PSA) is widely used to quantify risk by integrating accident frequencies and radiological consequences. However, the comprehensive PSA process, particularly level 3 consequence assessment using codes such as MACCS, demands significant computational resources and is not well-suited for rapid evaluations during design optimization. This challenge is intensified by recent Risk-Informed and Performance-Based (RIPB) regulatory frameworks, which require comprehensive risk quantification explicitly accounting for accident consequences across diverse accident sequences. To address this limitation, this study proposes a computationally efficient methodology based on a correlation model linking source-term releases to resulting radiological consequences. This approach enables the development of intuitive risk profiles that can enhance the practical utilization of risk information. As a case study, the proposed framework was applied to core damage accident sequences identified in the OPR-1000 level 1 PSA model, focusing on two initiating events: 1) loss of feedwater (LOFW), and 2) small break loss of coolant accident (SLOCA). Frequency-Consequence (F-C) curves were developed to compare the relative risks of the two events and evaluate the effects of severe-accident mitigation measures. The results demonstrate the initial applicability of the proposed methodology and indicate its potential to support more efficient risk-informed safety evaluations and decision-making.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106242"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927141","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 CHF with low mass flux in a full-scale rod based on the heated perimeter equivalent method","authors":"Binzhuo Xia ,&nbsp;Di Liu ,&nbsp;Kui Zhang ,&nbsp;Jinquan Yan ,&nbsp;Zhan Liu ,&nbsp;Fanting Xia ,&nbsp;Ronghua Chen ,&nbsp;Fujun Gan ,&nbsp;Wenxi Tian ,&nbsp;Bo Yang ,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.106219","DOIUrl":"10.1016/j.pnucene.2025.106219","url":null,"abstract":"<div><div>This study investigates the critical heat flux (CHF) phenomenon and the relevant parametric trends under low mass flux conditions, employing a single rod test section designed using the heated perimeter equivalent approach. Experimental results indicate that CHF increases with increasing mass flux and inlet subcooling, whereas its response to pressure exhibits a non-linear trend, initially rising and subsequently decreasing. Under zero mass flux conditions, CHF predominantly occurs in the middle and upper regions of the test section, accompanied by pronounced wall temperature fluctuations at low pressures. These findings provide valuable insights for nuclear reactor safety design, demonstrating that CHF values obtained from a heated perimeter equivalent single rod yield more conservative estimates compared to those derived from rod bundle configurations. Moreover, a novel methodology for identifying backflow was developed based on the backflow criterion number, which exceeds 90 in the turbulent regime and surpasses 400 in laminar and transitional flow regimes. Analysis of the experimental data confirms the effectiveness of the heated perimeter equivalent approach in accurately capturing the CHF behavior observed in real fuel assemblies. This research contributes to the improvement of CHF prediction models and enhances the fundamental understanding of CHF mechanisms under low mass flux conditions in nuclear thermal-hydraulic systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106219"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927225","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 dynamic characteristics of disturbance waves in a 3×3 rod bundle channel","authors":"Guangyuan Jin,&nbsp;Jinghu Bai,&nbsp;Weilian Li,&nbsp;Hongye He,&nbsp;Yandong Hou","doi":"10.1016/j.pnucene.2025.106211","DOIUrl":"10.1016/j.pnucene.2025.106211","url":null,"abstract":"<div><div>The dynamic characteristics of disturbance waves in a 3 × 3 rod bundle channel were investigated utilizing high-speed photography. The numerical simulations were performed to validate the experimental results, focusing on wave types, structural parameters and formation mechanisms. The disturbance wave structures were categorized into three types: small-scale waves, packet-like waves and band-like waves. The average liquid film thickness, wave height, and wave velocity exhibited a gradual increase with higher liquid superficial velocity. The effect of liquid velocity on wave frequency was primarily determined by the morphology of the disturbance waves. When liquid-phase velocity was held constant, the wave height decreased with increasing gas superficial velocity. Under identical flow conditions, the wave height gradually decreases from the corner rod to the edge rod and then to the center rod. The trends of wave velocity and wave frequency in relation to gas velocity and rod types were opposite to those of wave height. The velocity and pressure fields of the liquid phase responsible for generating the disturbance waves were analyzed. The generation of disturbance waves was driven by the high-speed turbulence of the gas phase and the wall reaction mechanism on the rod surface.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106211"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841094","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 fault diagnosis method for unknown faults in nuclear power plants considering domain adaptation and spatio-temporal feature fusion","authors":"Yushun Wang,&nbsp;Jingquan Liu","doi":"10.1016/j.pnucene.2025.106209","DOIUrl":"10.1016/j.pnucene.2025.106209","url":null,"abstract":"<div><div>The fault diagnosis in nuclear power plants presents the significant challenge of an open-set scenario, where fault types encountered during the deployment phase were not known during the training phase. However, current mainstream methodologies predominantly operate based on a closed-set assumption. This approach may result in misdiagnosis or nondiagnosis of unknown faults as some form of known fault, thereby misleading operators and yielding consequential verification outcomes. In this paper, we introduce a novel approach that integrates CNN, LSTM, OpenMax, and EVT to enhance fault diagnosis performance in open-set scenarios. Fault diagnosis in nuclear power plants can primarily be divided into system-level and component-level assessments. To validate the efficacy of our proposed method across these domains, we performed fault diagnosis on both bearing assemblies and pressurized water reactor systems. Additionally, we explored the impact of various factors on diagnostic accuracy, including distance metrics, tail size, and openness. Our experiments demonstrate that the proposed model delivers high accuracy in open-set fault diagnosis at nuclear power stations. Typically, with an openness of less than 10%, an accuracy exceeding 90% can be achieved. Furthermore, among the three evaluated measurement methods, the cosine-based method shows the highest accuracy and stability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106209"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841099","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}