Nuclear Engineering and Design最新文献

{"title":"A quantitative analysis of ATF surface characteristics on critical heat flux using Machine learning","authors":"Bruno P. Serrao ,&nbsp;Ye Kwon Huh ,&nbsp;Eliot Ciuperca ,&nbsp;Elvan Sahin ,&nbsp;Kaibo Liu ,&nbsp;Juliana P. Duarte","doi":"10.1016/j.nucengdes.2025.113924","DOIUrl":"10.1016/j.nucengdes.2025.113924","url":null,"abstract":"<div><div>The effects of surface characteristics on pool boiling Critical Heat Flux (CHF) are qualitatively understood based on previous investigations. However, more quantitative analyses are needed since the existing CHF correlations do not provide good predictions for modified surfaces. Using machine learning (ML) models as a tool, this study performed a quantitative analysis of relevant CHF parameters under pool boiling conditions: pressure, a dimensional feature, average roughness, static contact angle, surface orientation, and substrate thermal effusivity. A database was constructed by collecting accident tolerant fuels (ATF) CHF experimental data available from fourteen published studies. After hyperparameter optimization, the random forest (RF) model was selected for achieving the best fitting scores relative to other tested models. Feature importance models ranked static contact angle and pressure as the most important features, which is consistent with some of the literature CHF predictive models that take the surface characteristics into consideration. Finally, CHF predictions were obtained and compared to CHF experimental data. The effect of each feature on CHF was analyzed, while keeping other features fixed, by observing the experimental and predicted datapoints trends. The RF model demonstrated the ability to capture the experimental data trends, showing the RF model is suitable as a predictive pool boiling CHF model for this database.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113924"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143453643","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 safe disposal technology and progress of radioactive nuclear waste","authors":"Duo Dong ,&nbsp;Ziqi Wang ,&nbsp;Jingyu Guan ,&nbsp;Yi Xiao","doi":"10.1016/j.nucengdes.2025.113934","DOIUrl":"10.1016/j.nucengdes.2025.113934","url":null,"abstract":"<div><div>Nuclear energy as an important way to solve the energy crisis, has achieved multi-dimensional utilization in fields of power generation, heating, hydrogen production, seawater desalination, and isotope production. The safe disposal of radioactive waste directly affects the sustainable development of nuclear energy. The solidification of radioactive waste on specific substrates and deep geological disposal is an effective way to achieve permanent shielding of radioactive waste. The paper focuses on the latest developments in the disposal of radioactive waste, and comprehensively elaborating on various solidification disposal technologies for radioactive waste (cement solidification, asphalt solidification, polymer solidification, artificial rock solidification, glass solidification), as well as melting preparation processes (tank method, calcination + induction heating metal melting furnace, Joule heating ceramic melting furnace, cold crucible). In addition, the design principles, confinement mechanism, and application status of typical glass solidification technology were discussed, and the characteristics of typical borosilicate and phosphate glass matrices used to envelop radioactive nuclides were analyzed. Finally, the future development path of radioactive waste solidification disposal was pointed out, and providing reference for promoting the safe and efficient disposal of radioactive waste in the nuclear industry.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113934"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143453645","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 on-line structural integrity assessment system for the primary loop pipeline in NPPs","authors":"Mingya Chen ,&nbsp;Fangjie Shi ,&nbsp;Changjin Geng ,&nbsp;Weiwei Yu ,&nbsp;Quanjia Peng ,&nbsp;Wanxian Zhao ,&nbsp;Shuiyong Wang ,&nbsp;Guanbin Ma ,&nbsp;Yaolei Han","doi":"10.1016/j.nucengdes.2025.113908","DOIUrl":"10.1016/j.nucengdes.2025.113908","url":null,"abstract":"<div><div>A comprehensive online structural integrity assessment (SIA) methodology that fully accounts for the actual aging conditions and real-time transient operating parameters of primary loop pipelines in nuclear power plants remains absent in the existing literature. This paper proposes a novel online SIA approach based on thermoelectric potential (TEP) and failure assessment diagram (FAD) technologies. Initially, the online SIA process is established, and a conversion method between TEP values and impact energy is explored. The findings demonstrate that the Seebeck coefficient is highly sensitive to microstructural changes in the material, and the variation trend of TEP values closely aligns with that of impact energy. Subsequently, a systematic conversion methodology for TEP values and fracture toughness is presented. The FAD for primary loop pipelines is developed and analyzed. Sensitivity analysis of the safety margin reveals that enhancing the flow stress strength of the material or reducing the structural load is an effective strategy for improving the safety margin of the primary loop pipeline. Finally, an online SIA software, which has been successfully applied in the assessment of a nuclear power plant, is introduced. The SIA-based results confirm that the pipeline remains safe during the actual startup process.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113908"},"PeriodicalIF":1.9,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143453644","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":"Formation of gas bubble superlattice in U-Mo alloys: A phase-field study","authors":"Yanbo Jiang ,&nbsp;Shisen Gao ,&nbsp;Yongxiao La ,&nbsp;Wenbo Liu","doi":"10.1016/j.nucengdes.2025.113912","DOIUrl":"10.1016/j.nucengdes.2025.113912","url":null,"abstract":"<div><div>The impact of grain boundary (GB) on the formation and evolution of gas bubble superlattices (GBS) in U-Mo alloys under irradiation is critical for understanding the material behavior in nuclear environments. In this study, a phase-field model coupled Kim-Kim-Suzuki (KKS) model and explicit nucleation algorithm was developed to simulate GBS formation. The accumulation of vacancies and gas atoms led to bubble nucleation, with directional migration of interstitial atoms inducing a shadow effect and causing ordered bubble arrangements. The GBS exhibited stability, with bubble size and lattice constants remaining nearly constant at higher fission densities. The GB was shown to influence GBS formation significantly, with the surrounding region divided into a denuded zone and a peak zone. The width of the denuded zone is influenced by the GB properties. In this work, the relationship between the denuded zone width and the GB absorption strength was derived using a one-dimensional steady-state vacancy diffusion equation. It was found that the denuded zone width increases with an increase in the GB absorption coefficient. The phase-field simulation results were consistent with theoretical predictions. These findings contribute to a better understanding of how GBs affect irradiation-induced microstructural changes in nuclear materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113912"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444782","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":"Seismic performance analysis of suspended storage structure in lead-based reactor considering the coupled effects of liquid sloshing and soil condition","authors":"Yin Xunqiang,&nbsp;Chen Xu,&nbsp;Xu Shutong","doi":"10.1016/j.nucengdes.2025.113935","DOIUrl":"10.1016/j.nucengdes.2025.113935","url":null,"abstract":"<div><div>LEAd-based Reactor (LEAR) has good inherent safety, potentially making it the first commercial application of the fourth-generation nuclear power system. Facing the complex inland seismic geological environment, the sloshing of liquid heavy metal coolant with pool layout will significantly influence the seismic performance of suspended storage structure. To improve the seismic safety of the LEAR, a comprehensive study of the effect of liquid sloshing and soil conditions was discussed by combining the Compact Viscoelastic (CV) element for soil-structure interaction analysis, fluid–structure interaction for sloshing analysis and refined simulation of suspended storage structure within the framework of finite elements. And then, the verification example of cylindrical liquid storage vessel was presented to demonstrate the accuracy and efficiency of the proposed fluid–structure interaction model. Finally, the analysis of displacement, acceleration, maximum interlayer displacement angles, and other dynamic responses are carried out to study the effect of liquid sloshing features and soil conditions. The results showed that the properties of the coolant and internal components affect the response of the suspended storage structure, and the soft soil significantly increases the sloshing wave height of the internal fluid, while the overall displacement patterns and amplitude variations are relatively modest. The findings of this study may provide theoretical support and technical foundation for seismic safety evaluation and optimization design of LEAR suspended storage structure under the complex soil site.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113935"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143453642","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":"Buoyant-diffusive flow in the HTGR air ingress accident scenario","authors":"Zachary Welker ,&nbsp;Annalisa Manera ,&nbsp;Victor Petrov ,&nbsp;Paolo Balestra","doi":"10.1016/j.nucengdes.2025.113922","DOIUrl":"10.1016/j.nucengdes.2025.113922","url":null,"abstract":"<div><div>Experimental data from the Helium-Air Ingress gas-Reactor Experiment (HAIRE) has provided insight into an infrequently studied flow scenario where buoyant driven flow and molecular diffusion combine during an exchange flow. The combined buoyant-diffusive exchange flow occurs because of a transition region where buoyancy and molecular diffusion combine to create flow rates higher than the separate flow regimes would predict. The buoyo-diffusive flow is described by a dimensionless number which is used to quantitatively explain the results. The flow regime is of considerable importance in the small and medium-sized accident scenarios for High-Temperature Gas-cooled Reactors (HTGRs), where the flow regime will increase the air ingress rate in comparison to previously understood theory. For medium sized breaks the estimated increase in air ingress rate is up to 5% compared to the previous theory, and for small sized breaks the increase is either 5% or greater depending on the break size. This considerable increase in air ingress rates could affect the evolution of small- and medium-sized air ingress accidents and the overall damage to the HTGR’s graphite core. Buoyo-diffusive exchange flow could manifest in other areas where small physical scales and high mass diffusion are present.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113922"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444784","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":"Gen-IV pool reactors: Validation of sloshing CFD modelling and behavior under large seismic events","authors":"Vincent Moreau,&nbsp;Manuela Profir","doi":"10.1016/j.nucengdes.2025.113920","DOIUrl":"10.1016/j.nucengdes.2025.113920","url":null,"abstract":"<div><div>This paper addresses one aspect of the safety of compact lead-cooled pool fast reactors. Under a seismic event, there is concern that the surface sloshing could compromise the integrity of some structures or components. We investigate this issue by means of computational modelling and simulation using the STAR-CCM + commercial software.</div><div>In the first part, we simulate a partially filled vertical cylinder under harmonic forcing and quantitatively validate the modelling by comparison with experimental data of the free-surface height at different locations. Specifically, we capture with a very high precision a behavioral transition of the amplitude-frequency map near resonance.</div><div>In the second part, the modelling is applied to a fast reactor mock-up in water. It is validated again by comparison with experimental data under harmonic forcing. The numerical mock-up is then subjected to four seismic events of increasing intensity. The free-surface height and the forces applied to the mock-up vessel are monitored.</div><div>It is found that the forces applied to the vessel are strongly dominated by the highest frequency of the seismic acceleration. By applying a high filter cut at 1 Hz, the peak force is reduced by one order of magnitude but remains strongly related to the damped acceleration signal. Even then, sloshing forcing effects are observed but remain marginal.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113920"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444785","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":"Maximizing accuracy in severe accident simulations: An In-depth analysis of sampling methods for MELCOR code during station black-out in WWER-1000","authors":"R. Gharari ,&nbsp;R. Ahangari ,&nbsp;E. Hasanifard ,&nbsp;B. Roostaii","doi":"10.1016/j.nucengdes.2025.113925","DOIUrl":"10.1016/j.nucengdes.2025.113925","url":null,"abstract":"<div><div>This study explores the uncertainty in severe accident analysis of nuclear power plants, focusing on the thermal–hydraulic parameters of the WWER1000 reactor during station black-out scenarios. The primary challenge in such analyses lies in the inherent uncertainties associated with the simplifying assumptions embedded in computational codes. To address this, five prominent sampling methods—simple generation sampling, Latin hypercube sampling, SOBOL, HALTON, and LHS-SOBOL—are applied to improve the MELCOR code’s predictive accuracy. The results highlight that SGS produces non-uniform input distributions and performs poorly in comparison to the other methods. Additionally, key parameters, including hydrogen production, maximum containment pressure, and reactor pressure vessel rupture time, show significant variability, ranging from 150 to 602 kg, 14,100 to 28,000 s, and 0.62 to 1.05 MPa, respectively, with a 95 % confidence level. The findings emphasize the importance of selecting robust sampling techniques for enhancing the reliability of nuclear safety assessments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113925"},"PeriodicalIF":1.9,"publicationDate":"2025-02-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444783","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":"Visual experimental research on flow pattern transition mechanisms in vertical helically-coiled tube under liquid-solid-liquid coupled heat transfer conditions","authors":"Hengyuan Wang,&nbsp;Hailin Lei,&nbsp;Qiyu Xuan,&nbsp;Jiacheng Lou,&nbsp;Qisong Yang,&nbsp;Huixiong Li","doi":"10.1016/j.nucengdes.2025.113919","DOIUrl":"10.1016/j.nucengdes.2025.113919","url":null,"abstract":"<div><div>Helically-coiled tube steam generators are important equipment in nuclear power plants. The fluid on the primary side flows over the helically-coiled tube bundles, and exchanges heat with the fluid on the secondary side inside the helically-coiled tube steam generator. Therefore, the secondary side fluid in the helically-coiled tube is under a liquid-solid-liquid coupled heat transfer heating condition. Furthermore, the flow pattern characteristics and transition mechanisms of the secondary side fluid in a helically-coiled tube steam generator play a crucial role in improving the heat transfer efficiency of the generator and the secure and stable operation of nuclear power plants. Therefore, a visual experiment system of a helically-coiled tube steam generator was set up in this paper to observe and study the flow pattern characteristics and transition mechanisms of the secondary side fluid under liquid-solid-liquid coupled heat transfer condition. It was found that the flow patterns of the secondary side fluid include bubble flow, slug flow, stratified-wavy flow and stratified flow in helically-coiled tube. The convergence, contact, collision, coalescence of boiling bubbles and the growth of bubbles along the top of tube were the main flow pattern transition mechanisms under liquid-solid-liquid heat transfer condition. Additionally, the influence of parameters on flow pattern characteristics and transition mechanisms were analyzed, and it was concluded that an increase of heating water temperature in the primary side or an increase of mass flux of secondary side fluid promoted flow pattern transition. However, an increase of mass flux of heating water in the primary side had no obvious effect on the flow pattern transition. Finally, the flow pattern map and the prediction correlations of flow pattern transition boundary curves of the helically-coiled tube were established under liquid-solid-liquid coupled heat transfer conditions based on the experimental results. This flow pattern map was similar to the map obtained by Li (2018), Zhang and Chen (1983), Liu et al. (1976) and Li et al. (2024) based on the water–air experiment in the helically-coiled tube in terms of the shape of some flow pattern transition boundary curve. However, it was significantly different from the flow pattern maps obtained by Mandhane et al., 1974 in a horizontal tube, Mishima and Ishii, 1984 in a vertical tube and Barnea et al. (1980) in a inclined tube.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113919"},"PeriodicalIF":1.9,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444781","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 thermal conductivity prediction of dispersion nuclear fuel elements based on deep learning and property-oriented inverse design","authors":"Zekai Huang ,&nbsp;Yingxuan Dong ,&nbsp;Qida Liu ,&nbsp;Xiaoyu Hao ,&nbsp;Hong Zuo ,&nbsp;Qun Li","doi":"10.1016/j.nucengdes.2025.113918","DOIUrl":"10.1016/j.nucengdes.2025.113918","url":null,"abstract":"<div><div>The in-pile thermal performance of dispersion nuclear fuel elements is crucial for reactor safety. The uncertainty of the thermal conduction throughout dispersion fuel is primarily influenced by the nonuniform distribution of fuel particles in the meat, especially the agglomeration behavior of fuel particles. In this paper, a new method has been developed for the rapid and accurate prediction of the effective thermal conductivity (ETC) of dispersion nuclear fuel elements based on the deep learning method. A deep learning model is trained to establish an implicit correlation between the microstructure of dispersion nuclear fuel elements and their ETC, enabling a predictive model to estimate ETC from two-dimensional microstructural diagrams. The dataset is generated using the finite element method, which takes into account the nonuniform distribution characteristic of fuel particles. The microstructures which affect the decision-making of the predictive model are demonstrated by the saliency map. Based on the predictive model, an inverse design method of the distribution of fuel particles was conducted for the specific ETC by metaheuristic algorithms. This study confirms the feasibility of directly evaluating ETC from microstructural images and provides a property-oriented inverse design method for the meat microstructure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113918"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429687","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}