Progress in Nuclear Energy最新文献

{"title":"Evaluating local wind circulation metrics for radionuclide transport and dispersion: A practical approach for radiological safety","authors":"M.A. Hernández-Ceballos ,&nbsp;M. Sangiorgi ,&nbsp;N. Conte ,&nbsp;J.P. Bolívar","doi":"10.1016/j.pnucene.2026.106261","DOIUrl":"10.1016/j.pnucene.2026.106261","url":null,"abstract":"<div><div>Understanding the atmospheric dispersion and transport of radioactive materials is crucial for assessing radiological exposure and potential health risks, as well as for optimizing radiological environmental impact assessment and radiological monitoring networks. The dispersion of radionuclides following an accidental release from a nuclear facility is highly influenced by local wind circulation patterns, yet these effects are often overlooked in routine atmospheric dispersion assessments. This study evaluates the role of simple wind circulation indices—stagnation, recirculation, and ventilation—in shaping the dispersion of radioactive material, demonstrating their relevance for nuclear safety planning. The analysis focuses on the Almaraz Nuclear Power Plant (ANPP), where 1.256 atmospheric dispersion simulations were conducted using the RIMPUFF model over a four-year period (2012–2015) under different meteorological conditions. Considering the existing set of 84 monitoring stations included in the EURDEP system in an area of 200 km around the ANPP, the influence of each local atmospheric process is analyzed and characterized by taking the TGDR maximum values reached, and the number of monitoring stations affected in each simulation. On average, results demonstrate that high stagnation confines radionuclide plumes near the source, with maximum TGDR reaching 0.005 μSv/h and affecting up to 14 monitoring stations. In contrast, high recirculation enhances local accumulation, leading to, on average, peaks of 0.035 μSv/h and reducing the number of stations impacted (12 monitoring stations). High ventilation conditions promote wider dispersion, with maximum TGDR of 0.002 μSv/h affecting 10 monitoring stations. Extreme cases of each atmospheric process are also analyzed, showing distinct effects on the spatial distribution of affected monitoring stations. These findings highlight that wind circulation indices, derived from routine meteorological data, offer a straightforward yet effective means of anticipating dispersion behaviour in emergency scenarios.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106261"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189702","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":"Simulation study of the influence of the spacer grid on thermal-hydraulic characteristics in the rod bundle channel","authors":"Zuolong Wu ,&nbsp;Jiejin Cai ,&nbsp;Facheng Chen ,&nbsp;Heng Xie","doi":"10.1016/j.pnucene.2026.106287","DOIUrl":"10.1016/j.pnucene.2026.106287","url":null,"abstract":"<div><div>In order to study the influence of the spacer grid on thermal-hydraulic characteristics in a 4 × 4 rod bundle channel, a subcooled boiling multiphase flow model has been established, and the reliability of the numerical model has been verified. The calculated and experimental volume fraction of vapor and CHF are compared and are in good agreement. The CFD software STAR-CCM+ is used to study the influence of the spacer grid on the thermal-hydraulic characteristics, and to obtain the CHF and distributions of flow field in the rod bundle channel under multiple operating conditions. The results show that the spacer grid induces a CHF enhancement of more than 5.6%, an increase of the average secondary flow velocity maximum by more than 23 times, a reduction of maximum volume of vapor by more than 35.2%, and an increase in the total pressure drop in the rod bundle channel by more than 0.7 times. The dimple and spring structure dominate the pressure drop contribution, while the mixing vane structure performs better in secondary flow enhancement and lateral churning.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106287"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189700","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":"Analyzing the sources of radioactive waste from Small Modular Reactors (SMR)","authors":"Mumtaz Khan ,&nbsp;Hu Penghua ,&nbsp;Jie Niu ,&nbsp;Xu Lechang ,&nbsp;Wang Yalan ,&nbsp;Thaqal M. Alhuzaymi","doi":"10.1016/j.pnucene.2026.106239","DOIUrl":"10.1016/j.pnucene.2026.106239","url":null,"abstract":"<div><div>This review critically compares waste generation and back-end challenges across major SMR families—iPWR (e.g., NuScale), PHWR/CANDU-class, HTGR (Xe-100), SFR (e.g., ARC/Natrium-class), and MSR (e.g., IMSR)—against a large PWR baseline using repository-relevant metrics (package-scale decay heat at 1/10/100 years, mobile fission-product inventories/speciation, fissile content/criticality margins, and secondary-waste volumes). Quantitatively, independent assessments indicate that several SMR concepts can increase total waste requiring management by ∼2–30 × per MWh versus gigawatt-scale LWRs due to higher neutron leakage and reactive coolants; a representative iPWR can discharge ∼1.7 × more SNF per unit electricity than a large PWR, with additional activated-metal ILW from compact integral vessels. By contrast, the Xe-100 achieves very high burnup, yielding ∼75 % lower SNF mass and 35–50 % lower 100-year decay heat, but at the cost of ∼10–12 × larger geometric SNF volume dominated by GTCC graphite. SFRs may reduce SNF mass per MWh but produce TRU-rich, sodium-bonded fuel that typically requires pyroprocessing and generates distinct sodium-cleanup residues. CANDU-class SMRs lower long-lived TRU inventories yet increase SF package counts (low burnup) and require first-class tritium systems. MSRs/IMSR shift HLW to fuel-bearing salts and off-gas media, making salt-matched waste-form qualification (e.g., glass-bonded ceramics) decisive. We synthesize design-specific waste drivers, identify data gaps (e.g., FOAK→NOAK, per-MWh benchmarks), and outline research directions: (i) leakage/activation reduction, (ii) graphite and tritium mass-balance management, (iii) sodium neutralization and ER-salt waste forms, and (iv) qualification of salt-specific immobilization routes. Overall, the evidence favors evaluating SMRs on repository-facing metrics rather than bulk mass claims, with several concepts likely increasing per-MWh waste burdens unless back-end co-design is demonstrably effective.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106239"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981518","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":"Modeling flow quality for predicting axial void fraction development of boiling flows in rectangular channels at elevated pressure conditions","authors":"Shichang Dong,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106255","DOIUrl":"10.1016/j.pnucene.2026.106255","url":null,"abstract":"<div><div>Accurate estimation of the void fraction plays a critical role in the heat and mass transfer analysis in boiling flows within rectangular channels. The existing classic void fraction model utilizes the point of onset of significant void (OSV) as the beginning location for the axial void fraction development, which limits its ability to accurately predict void fractions around the OSV point. To address this limitation, a flow quality correlation for subcooled boiling flows was first developed and then extended to cover the entire boiling regime. Subsequently, an axial void fraction model for boiling flows in rectangular channels was formulated by incorporating a full-range drift-flux correlation. The model was rigorously validated using an extensive experimental database compiled from eight independent sources. Results show that, compared to classic models, the proposed model achieves more accurate predictions of axial void fraction evolution both upstream and downstream of the OSV point. Moreover, it demonstrates improved capability in predicting thermal equilibrium quality at the OSV point relative to the Saha-Zuber correlation. Additionally, the model accurately captures the influence of several key parameters on the void fraction profiles, indicating that this model is well applicable for boiling flow analysis in rectangular channels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106255"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039019","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":"Comparative analysis of the ACME and APEX thermal-hydraulic test facilities for advanced passive PWRs","authors":"Chengcheng Deng,&nbsp;Chengzhi Li,&nbsp;Junxiao Yang,&nbsp;Zongyang Li","doi":"10.1016/j.pnucene.2026.106266","DOIUrl":"10.1016/j.pnucene.2026.106266","url":null,"abstract":"<div><div>For the safety assessment of advanced nuclear reactors, it is necessary to design and construct scaled thermal-hydraulic test facilities to replicate key phenomena and processes of the prototype reactors. The Advanced Core-cooling Mechanism Experiment (ACME) and Advanced Plant Experiment (APEX) are two typical integral thermal-hydraulic test facilities designed and constructed for advanced passive PWRs. The APEX test facility was constructed for AP1000, while the ACME test facility was built for CAP1400 in China and selected as the benchmark facility for the International Standard Problem No.51 (ISP-51) project. In this study, comparative analysis between the ACME and APEX test facilities was conducted from both qualitative and quantitative perspectives. On one hand, the variation curves of key parameters during the Small Break Loss-of-Coolant Accident (SBLOCA) transient process were compared and analyzed by combining experimental data and simulation results of ACME and APEX. On the other hand, a system-level scaling analysis method was employed to quantitatively compare the dimensionless numbers of key phenomena in different stages during SBLOCA transient. The results indicate that both the ACME and APEX are well-scaled thermal-hydraulic test facilities for examining the behavior of passive safety systems under the SBLOCA transient process. Moreover, the ACME and APEX test facilities exhibit good similarity during the mid-to-late stages of the SBLOCA transient process. Through the comparative analysis of ACME and APEX facilities, this study can provide guidance for the scaling design and interactive verification of experimental data of integral thermal-hydraulic test facilities designed for advanced nuclear reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106266"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039018","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 on pressure oscillation characteristics of unstable steam jet condensation through round/obround side-hole spargers","authors":"Jiayu Xiao,&nbsp;Zhongning Sun,&nbsp;Haozhi Bian,&nbsp;Zhaoming Meng","doi":"10.1016/j.pnucene.2025.106202","DOIUrl":"10.1016/j.pnucene.2025.106202","url":null,"abstract":"<div><div>Unstable steam jet condensation may cause strong pressure oscillation. Transient numerical simulation is conducted on unstable steam jet for round and obround side-hole spargers. The model effectively captures differences in main frequency of the two type holes. The relationship between pressure fluctuation characteristics in fluid domain and bubble evolution process is discussed. The pressure peak occurs near the secondary bubble collapse location. Under the same thermal conditions, obround holes produce a time-averaged bubble volume that is 1.22 times larger, a time-averaged penetration depth that is 0.96 times that of round holes, and condensation oscillation periods that are 1.16 times longer. The pressure distribution in fluid domain is reconstructed through data of 112 pressure monitoring points. A pressure peak area exists in the fluid domain, and the magnitude of pressure peak at each position in domain is almost inversely proportional to distance from position where the secondary bubble collapses. Position of the maximum pressure peak on sparger surface occurs at a distance of 0.5–1.5 times diameter from the hole edge. With increase of steam mass flux and water subcooling, the maximum peak pressure in domain increases, meanwhile, the pressure impact on sparger wall also increases.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106202"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981520","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":"Modeling fuel behavior in liquid metal fast reactors: A multiphysics approach with JOG formation analysis","authors":"Mou Wang ,&nbsp;Gen Jiang ,&nbsp;Kai Wang ,&nbsp;Songbai Cheng ,&nbsp;Wenzhong Zhou","doi":"10.1016/j.pnucene.2026.106276","DOIUrl":"10.1016/j.pnucene.2026.106276","url":null,"abstract":"<div><div>The long-term performance evolution and fission gas release (FGR) behavior of liquid metal-cooled fast reactor (LMFR) fuel elements are crucial for reactor safety and radioactive source term assessment. In this paper, to address the deficiencies of fuel performance analysis models for LMFR, a multi-physics field-coupled fuel performance analysis program is developed by using the FAST and CAMPUS program architectures and integrating the key physics models of FEAST, KMC-fuel, and other fast reactor programs. The program contains core modules for thermal-physical analysis, FGR, and chemical element migration (oxygen/plutonium), and considers a joint oxide gain (JOG) formation module. The program adopts two-dimensional axisymmetric geometry modeling to enhance the computational efficiency. Based on the verification of the program simulation results in comparison with the irradiated data of the experimental reactor, the present model shows high accuracy in the prediction of fuel temperature field distribution, gap closure kinetics, and fission gas release share (average relative error with experimental data is significantly lower than that of the FEAST model), and that the overall program is able to simulate the evolution of the overall performance of the fuel element well. Based on the validation of the overall performance of the program, the study also analyzes the key role of line power and JOG formation on fuel performance. It is shown that the pellet thermodynamic temperature rises significantly with increasing fuel operating power, which exacerbates the FGR behavior and induces an increase in cladding stress. More critically, the formation of JOG enhances the thermal conductivity of the fuel gap, which changes the temperature field distribution of the pellet, the mechanical deformation of the material, etc., and is thus a key factor that should not be ignored in the process of accurately predicting the change in fuel performance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106276"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146079638","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":"UO2 oxidation in dry storage conditions: from new data to exploratory modelling of consequences","authors":"A. Milena-Pérez,&nbsp;F. Feria,&nbsp;C. Aguado,&nbsp;S. Fernández-Carretero,&nbsp;L. Gutiérrez,&nbsp;N. Rodríguez-Villagra,&nbsp;H. Galán,&nbsp;L.E. Herranz","doi":"10.1016/j.pnucene.2026.106288","DOIUrl":"10.1016/j.pnucene.2026.106288","url":null,"abstract":"<div><div>In this work, we have studied the oxidation behavior of a fresh UO₂ surrogate in representative conditions of a dry interim storage for SNF, both in terms of temperature (200–400 °C) and oxygen partial pressure (0.1–21% O₂). The reaction has been studied <em>in-situ</em> by TGA and <em>ex-situ</em> by autoclaves. XRD and Rietveld's refinement has been used for quantification of uranium oxidized phases, with particular interest in U₃O₈. The predominance of temperature over oxygen concentration in the formation of U₃O₈ has been confirmed. TGA results have shown that, at 200 and 250 °C, no U₃O₈ has been detected after 10 h of thermal treatment, not even in air (21% O₂). Only at 275 °C, U₃O₈ has been detected at 21 and 1% O₂. This result allows better defining the temperature threshold of U₃O₈ formation in unirradiated UO₂. On the upper limit, 350 and 400 °C, the reaction proceeds up to quantitative formation (<em>i.e.,</em> more than 90%) of U₃O₈, even with very low oxygen content (as low as 0.1% O₂). Finally, at intermediate temperatures (<em>i.e.,</em> 300 °C), oxygen partial pressure starts playing a role, being the extent of the oxidation lower with lower oxygen concentration. This effect has been seen when studying the reaction at longer times by using <em>ex-situ</em> autoclaves. However, given enough time, all the fuels end showing a majority of U₃O₈, which is between 6 and 24 h for 10 and 1% O₂ and between 24 h and 1 week for 0.1% O₂. In all the experiments, 21 and 10% O₂ behave in a very similar way for all the temperatures, but there is a substantial change when decreasing oxygen to 1% O₂, being in this case the oxidation much more limited. This could be an indication of an oxygen concentration threshold that could trigger the nucleation and growth of U₃O₈ in a greater extent. Based on the experimental work and previous modelling studies, an analytical methodology to assess the cladding integrity during UO₂ oxidation has been developed by adapting FRAPCON-xt, a home-extended fuel performance code to dry storage.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106288"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189226","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":"Palladium corrosion of the SiC layer in TRISO particle fuels: A separate effects study to assess mechanisms for palladium attack","authors":"Steven J. Cavazos ,&nbsp;Julian Valdez ,&nbsp;Jordan Stone ,&nbsp;Carlos Hernandez ,&nbsp;Da Teng ,&nbsp;Gisele George ,&nbsp;Brian D. Wirth ,&nbsp;Elizabeth Sooby","doi":"10.1016/j.pnucene.2026.106290","DOIUrl":"10.1016/j.pnucene.2026.106290","url":null,"abstract":"<div><div>Palladium (Pd) is a fission product of particular interest for TRISO fuel performance due to its propensity to chemically corrode the SiC layer. Post irradiation examination experiments on irradiated TRISO particles confirm the formation of several palladium silicide intermetallics, such as PdSi and Pd₂Si, in the SiC layer. A corroded SiC layer can lead to TRISO particle failure, resulting in higher fission product release. However, the specific transport mechanisms driving Pd corrosion remain difficult to isolate during irradiation tests. Therefore, this study aims to quantify the thermal diffusion kinetics of Pd through the OPyC layer and its subsequent interaction with the SiC layer under separate effects conditions. The research presented uses surrogate TRISO particles, with Pd-rich coatings on the OPyC layer, and high temperature separate effects testing to examine the behavior of Pd diffusion and chemical interactions on the TRISO particle layers. Physical vapor deposition techniques were used to deposit 200 nm of Pd on surrogate TRISO particles, which were then annealed at various high temperatures. SEM/EDS characterization was used to evaluate the annealed microstructures and revealed a distinct thermal activation threshold. While no Pd diffusion was observed at 850 °C after 10 h, annealing at 1150 °C resulted in complete diffusion through the OPyC layer and Pd-SiC interaction along the OPyC-SiC interface. Pd diffusion was accompanied by OPyC degradation, including OPyC cracking and Si migration to the particle surface. These findings establish 850–1150 °C as a critical transition regime for Pd chemical attack, providing experimentally derived reaction limits necessary for validating fuel performance models and defining safety margins for high-temperature transients. This research characterizes thermally driven Pd diffusion in TRISO fuel layers under non-irradiated conditions, contributing to improved predictive capability for fuel integrity in advanced nuclear energy systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106290"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189698","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":"Study on the interface characteristics of bubble condensation with non-condensable gas","authors":"Cun Liu ,&nbsp;Liangming Pan ,&nbsp;Longxiang Zhu ,&nbsp;Jiewen Deng","doi":"10.1016/j.pnucene.2026.106273","DOIUrl":"10.1016/j.pnucene.2026.106273","url":null,"abstract":"<div><div>Bubble condensation, as a form of direct contact condensation, is widely employed in engineering applications due to its high heat transfer efficiency. However, current understanding of the condensation behavior of bubbles containing non-condensable gases remains limited, particularly regarding the mechanism by which non-condensable gases impede bubble heat and mass transfer. In this study, the Volume of Fluid (VOF) method coupled with the Lee model and a species transport model is used to numerically simulate the condensation process of vapor bubbles containing non-condensable gas. By systematically varying key parameters—including initial bubble volume, liquid subcooling, and vapor mass fraction—the influence of non-condensable gas on bubble heat and mass transfer behavior is investigated. Results indicate significant differences in condensation rate and morphological evolution between pure vapor bubbles and those containing non-condensable gas. During the condensation of bubbles with non-condensable gas, the initial condensation rate shows a weak linear correlation with vapor mass fraction, whereas a strong linear relationship emerges in later stages. The content of non-condensable gas notably affects bubble shape evolution and rise velocity. Specifically, condensation initially concentrates at the lateral surface of the bubble, then gradually shifts toward the top and bottom in later stages. Moreover, under conditions of higher subcooling and higher vapor mass fraction, the internal flow within the bubble exhibits distinct disordered characteristics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106273"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146079637","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}