Annals of Nuclear Energy最新文献

{"title":"Thermo-mechanical performance of petal-shaped fuel rods based on accident tolerant fuel cladding","authors":"Binxian He ,&nbsp;Juexiong Deng ,&nbsp;Wenchao Zhang ,&nbsp;Xiangfei Meng ,&nbsp;Jianchuang Sun ,&nbsp;Yuxiang Hong ,&nbsp;Weihua Cai","doi":"10.1016/j.anucene.2025.111863","DOIUrl":"10.1016/j.anucene.2025.111863","url":null,"abstract":"<div><div>Accident-tolerant fuel (ATF) cladding enhances nuclear fuel performance under extreme accident conditions. Current research on applying ATF cladding to petal-shaped fuel rods (PSFRs) for performance improvement remains exploratory. This study establishes a finite element-based thermo-mechanical coupling model to analyse PSFR behaviour under irradiation. The thermal–mechanical performance of Cr-coated, FeCrAl-coated, Zr-1%Nb, and FeCrAl claddings was investigated under normal operation and reactivity-initiated accident (RIA) conditions to evaluate their potential for replacing conventional Zr-4 cladding. Results indicate that thermal creep is the primary factor driving performance differences among cladding materials. Cr/FeCrAl-coated PSFR exhibited minor deviations from the reference model in stress, strain, and temperature distribution. While FeCrAl cladding demonstrated superior deformation resistance, it displayed earlier plastic strain initiation. Zr-1%Nb achieved significantly lower cladding stress (15.6 MPa at 5 % FIMA) owing to its excellent creep rate, highlighting its advantage in stress management. However, creep-induced radial deformation necessitates either increasing the gap between fuel rods or localizing the cladding thickness at petal angle to mitigate dimensional impacts.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111863"},"PeriodicalIF":2.3,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145098675","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":"Thermal resistance evolution of NiFe2O4 particle deposition on nuclear fuel rods during subcooled boiling heat transfer","authors":"Yu Zhao,&nbsp;Zhizhong Tan,&nbsp;Jian Zheng","doi":"10.1016/j.anucene.2025.111881","DOIUrl":"10.1016/j.anucene.2025.111881","url":null,"abstract":"<div><div>The accumulation of particulate fouling deposition on nuclear fuel rod surfaces critically impacts reactor safety, operational stability, and economic efficiency. This study presents a numerical investigation of NiFe₂O₄ particulate deposition on nuclear fuel rods. The framework integrates ANSYS FLUENT’s Discrete Phase Model (DPM) with customized deposition algorithms to resolve particle-laden flow dynamics and fouling mechanisms. Additionally, Boiling surface deposition dynamics were rigorously investigated through coupled Lee phase-change modeling and three-dimensional VOF interfacial tracking. Numerical results exhibit excellent agreement with experimental measurements, validating the model’s predictive accuracy. NiFe₂O₄ particulate fouling exhibits significantly higher deposition propensity under pressurized water reactor (PWR) conditions. Parametric analysis reveals accelerated convergence to equilibrium fouling resistance with elevated flow velocities and larger particle diameters. The fouling resistance asymptotic value directly correlates with flow rate reduction and particle size increase. Boiling-induced vapor bubble nucleation suppresses deposition efficiency through microconvection effects, extending stabilization periods compared to non-boiling conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111881"},"PeriodicalIF":2.3,"publicationDate":"2025-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145098671","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 hydrodynamic loads during air clearing process of safety relief system for NHR200-II reactor","authors":"Li shu ,&nbsp;Zhang Dandi ,&nbsp;Tong Lili ,&nbsp;Cao Xuewu","doi":"10.1016/j.anucene.2025.111871","DOIUrl":"10.1016/j.anucene.2025.111871","url":null,"abstract":"<div><div>In response to the design requirements of the safety discharge system for the primary loop of NHR200-II small onshore reactor, an experimental simulation device for the transient discharge system was established, with full pressure and flow rate during the air clearing process. Experimental study on the transient dynamic loading behavior of the water tank were conducted under the initial discharge pressure of 10 MPa. The influences of the submergence of sparger, pool temperature, discharge gas temperature, non-condensable gas fraction, and valve opening time on the amplitude and main frequency of dynamic load oscillation were obtained. The dynamic pressure amplitude under all conditions is between 23.8 kPa and 48.7 kPa, with a main frequency between 2.3 Hz to 3.1 Hz, showing low frequency and high amplitude characteristics. The findings indicated that the oscillation characteristics were most significantly influenced by submergence depth of the bubbler and the duration of valve opening, followed by water temperature, non-condensable gas fraction and discharge gas temperature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111871"},"PeriodicalIF":2.3,"publicationDate":"2025-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145057201","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":"The feasibility of calculating the enrichment of 235U using the PIGE method with high-energy protons","authors":"Marzieh Hassanpour ,&nbsp;Mehdi Hassanpour ,&nbsp;MohammadReza Rezaie ,&nbsp;Mohammad Rashed Iqbal Faruque","doi":"10.1016/j.anucene.2025.111882","DOIUrl":"10.1016/j.anucene.2025.111882","url":null,"abstract":"<div><div>In this study, an attempt has been made to investigate the ²³⁵U enrichment percentage in uranium compounds using the PIGE method with 25 MeV proton energy and to calculate the characteristic gamma energy of ²³⁵U. The results indicate that the characteristic gamma energy of 2.2 MeV corresponds to ²³⁵U in the uranium compound, where no gamma line from ²³⁸U is observed at this energy. By altering the percentage of ²³⁵U enrichment in the compound, the response of the CsI(Tl) detector to the 2.2 MeV gamma is formulated as a function of different enrichment percentages, utilizing the results obtained from MCNPX simulations. Furthermore, by examining various nuclear processes resulting from the interaction of protons with ²³⁵U and ²³⁸U, the reason for the gamma radiation at 2.2 MeV for ²³⁵U is also elucidated.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111882"},"PeriodicalIF":2.3,"publicationDate":"2025-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045215","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":"High-parameter multiphase critical flow in supercritical carbon dioxide nuclear energy system","authors":"Gengyuan Tian ,&nbsp;Yuan Zhou ,&nbsp;Yanping Huang ,&nbsp;Yuan Yuan ,&nbsp;Chengtian Zeng ,&nbsp;Shuai Liu","doi":"10.1016/j.anucene.2025.111878","DOIUrl":"10.1016/j.anucene.2025.111878","url":null,"abstract":"<div><div>The loss-of-coolant accident (LOCA) is one of the major safety concerns in the supercritical carbon dioxide power cycle system. The precise calculation of critical flow rate during LOCA is of paramount significance as it determines the rate of coolant loss in the system. In this paper, a set of simple regenerative power cycle experimental apparatus was established, and research on high-parameter (maximum pressure:15.0 MPa, maximum temperature:500.0 °C) critical flow was carried out. The influence of thermal parameters and geometric parameters on critical flow rate was analyzed. A theoretical model of critical flow that accounts for thermal non − equilibrium phase transition and variation of the compressibility has been developed. The model accurately predicts supercritical CO₂ critical flow rates across wide parameters, with under 10 % error. The experimental data and theoretical model obtained in this research can be utilized for the safety analysis of supercritical CO₂ power cycle systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111878"},"PeriodicalIF":2.3,"publicationDate":"2025-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145057200","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":"Typical application studies of thermal management technology in nuclear reactor","authors":"Youyou Xu,&nbsp;Jian Deng,&nbsp;Dahuan Zhu,&nbsp;Qi Lu,&nbsp;Jia Liu,&nbsp;Wei Zeng,&nbsp;Xu Ran,&nbsp;Zhifang Qiu,&nbsp;Dongwei Wang","doi":"10.1016/j.anucene.2025.111861","DOIUrl":"10.1016/j.anucene.2025.111861","url":null,"abstract":"<div><div>Thermal management technology is an indispensable key technology in the development of modern high-precision electronic devices, aiming to control the temperature of target objects within an allowable range. In nuclear reactors, characterized by high power density and long operating duration, represent a sustainable and clean energy source, favored by countries around the world. During normal operation, it is necessary to remove heat from the reactor core and heat-generating equipment to ensure the core operates safely and stably; in accident scenarios, the core of accident response failure in removing residual heat from the core to prevent temperatures from exceeding safety limits and causing more severe consequences. As the most complex system in a reactor, researching efficient thermal management systems is a necessary condition for ensuring the safety and efficient operation of DCS(Distributed control system) systems. DCS systems require the development of efficient thermal management systems to ensure their safe and efficient operation. The success or failure of thermal management in DCS systems directly impacts their performance, reliability, environmental comfort, and energy efficiency. In the industrial sector, thermal management technology has evolved from traditional air-cooling to more efficient cooling methods such as liquid cooling and phase-change cooling, significantly improving thermal management efficiency while enabling miniaturization and compactness of thermal management systems. Currently, DCS systems typically employ air-cooling for heat dissipation, resulting in large system sizes, high operational noise, and poor temperature control performance. With the increasing demands on DCS systems for miniaturization, modularization, digitization, and intelligence in nuclear power, coupled with the significant improvement in chip performance leading to a substantial increase in heat flux density, the air-cooling method with its low efficiency has become inadequate to meet the higher heat dissipation requirements of DCS systems. This paper conducts research on the application of thermal management technology in DCS thermal control design. A modular DCS board module model is established, and both air-cooling and liquid loop cooling methods are modeled, calculated, analyzed, and compared. Sensitivity analyses are also performed to provide references for the next step in designing efficient thermal management systems for DCS systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111861"},"PeriodicalIF":2.3,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045214","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 reactor refueling optimization using KAADPN integrated probability distribution guided heuristic algorithm","authors":"Yanpeng Sun,&nbsp;Xubo Ma","doi":"10.1016/j.anucene.2025.111862","DOIUrl":"10.1016/j.anucene.2025.111862","url":null,"abstract":"<div><div>This study addresses the refueling optimization problem for reactors, selecting the effective multiplication factor as the metric for evaluating loading schemes. The Characteristic Statistical Simulated Annealing and Characteristic Statistical Genetic Algorithm are proposed, which significantly enhance the exploration of the solution space and improve the global search capability. The Kolmogorov-Arnold Attention Dual-Path Network (KAADPN) is introduced, combining the function modeling ability of KAN with the global feature capture of the self-attention mechanism. This significantly improves the model’s prediction accuracy while enhancing its computational efficiency. By establishing a surrogate model for core physics calculations and integrating it with optimization algorithms, pseudo-equilibrium optimization analysis is conducted. The effectiveness of the algorithms is compared through single-cycle optimization case studies, and preliminary no-shuffling optimization verification is performed, resulting in ideal core fuel loading schemes. This validates the feasibility of the method and provides a new tool for efficiently addressing the refueling optimization problem.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111862"},"PeriodicalIF":2.3,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045212","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":"Bayesian network-based surrogate model for physical simulation in probabilistic risk assessment of nuclear power plants","authors":"Jingyu Chen,&nbsp;Tatsuya Sakurahara","doi":"10.1016/j.anucene.2025.111869","DOIUrl":"10.1016/j.anucene.2025.111869","url":null,"abstract":"<div><div>In recent years, physical simulations, such as computational fluid dynamics (CFD) and finite element analysis (FEA), have been increasingly utilized to support probabilistic risk assessment (PRA) of nuclear power plants (NPPs). However, these simulations are often computationally intensive, particularly when used to explore broad scenario spaces and quantify uncertainty in PRA. This paper proposes a Bayesian network (BN)-based surrogate modeling approach to reduce the computational burden of physical simulations in PRA. Methodological steps for developing, validating, and applying the BN-based surrogate model (specifically, to support screening analysis in PRA) are presented. The implementation of the proposed approach is demonstrated through a case study of NPP fire modeling, where a BN-based surrogate model is constructed and validated for the Consolidated Model of Fire Growth and Smoke Transport (CFAST) code. Compared to other machine learning-based surrogate models previously studied in the PRA domain, BNs offer two key advantages: (i) representing causal relationships among variables explicitly, and (ii) providing a transparent model structure and inputs that can be effectively communicated using diagrams and tables. Additionally, the widespread adoption of BNs in the PRA domain can facilitate broader applications of the proposed approach.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111869"},"PeriodicalIF":2.3,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045211","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 study of bistable flow phenomena in square tube bundles based on Large Eddy Simulation","authors":"Junshuai Sun ,&nbsp;Rulei Sun ,&nbsp;Pengrui Qiao ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.anucene.2025.111868","DOIUrl":"10.1016/j.anucene.2025.111868","url":null,"abstract":"<div><div>The nuclear energy industry is growing in the pursuit of high efficiency and safe operation. Consequently, the performance demand of key equipment, such as steam generators, is increasing. The existence of the bistable flow phenomenon in the tube bundles can result in a dynamic change in the fluid forces acting on the tubes. This, in turn, can have a non-negligible influence on the flow-induced vibration behaviour of the tube bundles. To explore the mechanism of the bistable flow phenomenon on the fluid forces and provide reliable technical support for the analysis of flow-induced vibration of tube bundles, this paper has numerically studied the bistable flow phenomenon in a square tube bundle with <em>P</em>/<em>D</em> = 1.40 (<em>P</em>, tube pitch, <em>D</em>, tube diameter) by using Large Eddy Simulation (LES). The three-dimensional turbulence effect in the tube bundles can be effectively simulated when the tube spreading length is larger than <em>πD</em>/4. The two quasi-steady flow modes in the bistable flow phenomenon were captured and distinguished, and exhibited random switching characteristics, indicating a chaotic phenomenon. By analysing the flow characteristics of the different modes in detail, it was found that there were significant differences in the velocity and pressure distributions in terms of values and trends. An in-depth study was carried out on the fluid forces acting on the tubes, particularly the lift force. The influence of three physical mechanisms on the lift force was revealed: the shear layer shedding of the tube itself, the intermittent interference of the shear layer of the upstream tube, and the inter-tube vortex interactions. The Strouhal numbers (Sr) corresponding to these three mechanisms are 0.205, 0.296, and 0.387, respectively. These values provide key parameters for quantitatively analysing the frequencies of the three mechanisms, and in turn, provide an important theoretical basis for the analysis of the flow-induced vibration and optimization design of the tube bundles.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111868"},"PeriodicalIF":2.3,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045213","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":"An improved spatial-dependent model of neutron multiplicity counting for large-volume plutonium solution","authors":"Jia-Cheng Wang ,&nbsp;Xiao-Dong Huo ,&nbsp;Hai-Feng Yang ,&nbsp;Zeng Shao ,&nbsp;Kan Wang","doi":"10.1016/j.anucene.2025.111858","DOIUrl":"10.1016/j.anucene.2025.111858","url":null,"abstract":"<div><div>Neutron multiplicity counting is a non-destructive, passive technique for monitoring plutonium inventory. However, when extending the application from plutonium metal/plutonium oxides to large-volume plutonium solution systems, the original “point model” reveals significant shortcomings. Currently, while two types of improvements for the original “point model” has been proposed: (a) improved “point model” suitable for small-volume solution systems and (b) volume-weighted “point model” correcting for spatial dependence of solid systems, neither is suitable for large-volume solution systems. Based on the improved “point model”, we firstly employ the volume-weighted approach to derive a volume-weighted model suitable for solution systems, which however neglects the disparity between the induced fission source distribution and the initial source distribution in our opinion. Furthermore, by additionally incorporating the distribution of induced fission reactions as a weighting factor for the spatial dependence correction, we propose the composite-weighted model. Comparative analysis of simulation results from the improved “point model”, volume-weighted model, and composite-weighted model demonstrates that the composite-weighted model has the best performance, offering superior universality and accuracy, thereby confirming the necessity and validity of the improvements. Theoretically, the methodology for addressing spatial dependence in large-volume solution systems introduced in this study can also be used for other large-volume plutonium-containing material systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111858"},"PeriodicalIF":2.3,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145026713","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}