Progress in Nuclear Energy最新文献

Modeling fuel behavior in liquid metal fast reactors: A multiphysics approach with JOG formation analysis 液态金属快堆燃料行为建模：基于JOG形成分析的多物理场方法

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-28 DOI: 10.1016/j.pnucene.2026.106276

Mou Wang , Gen Jiang , Kai Wang , Songbai Cheng , Wenzhong Zhou

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.

液态金属冷却快堆（LMFR）燃料元件的长期性能演变和裂变气体释放（FGR）行为对反应堆安全和放射源期评估至关重要。本文针对小中子反应堆燃料性能分析模型的不足，利用FAST和CAMPUS程序架构，集成FEAST、KMC-fuel等快堆项目的关键物理模型，开发了多物理场耦合燃料性能分析程序。该程序包含热物理分析、FGR和化学元素迁移（氧/钚）的核心模块，并考虑联合氧化增益（JOG）形成模块。程序采用二维轴对称几何建模，提高了计算效率。将程序模拟结果与实验堆辐照数据进行对比验证，该模型在预测燃料温度场分布、间隙闭合动力学和裂变气体释放份额方面具有较高的准确性（与实验数据的平均相对误差显著低于FEAST模型），总体程序能够较好地模拟燃料元件整体性能的演变。在验证该方案整体性能的基础上，分析了线路功率和JOG形成对燃油性能的关键作用。结果表明，随着燃料运行功率的增加，球团的热力学温度显著升高，这加剧了FGR行为，导致包层应力增大。更为关键的是，JOG的形成增强了燃料间隙的导热性，从而改变了颗粒的温度场分布、材料的力学变形等，是准确预测燃料性能变化过程中不可忽视的关键因素。

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引用次数: 0

Study on the interface characteristics of bubble condensation with non-condensable gas 气泡冷凝与不可冷凝气体界面特性研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.pnucene.2026.106273

Cun Liu , Liangming Pan , Longxiang Zhu , Jiewen Deng

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.

气泡冷凝作为直接接触冷凝的一种形式，由于其具有较高的传热效率，在工程上得到了广泛的应用。然而，目前对含有不可冷凝气体的气泡的冷凝行为的理解仍然有限，特别是关于不可冷凝气体阻碍气泡传热和传质的机制。本文采用流体体积法（Volume of Fluid， VOF）与Lee模型和物种输运模型相结合的方法，对含不可凝气体的蒸汽泡的凝结过程进行了数值模拟。通过系统地改变初始气泡体积、液体过冷度和蒸汽质量分数等关键参数，研究了不凝性气体对气泡传热传质行为的影响。结果表明，纯汽泡和含不凝气体的汽泡在凝结速率和形态演化上存在显著差异。在气泡与非可凝气体的冷凝过程中，初始冷凝速率与水蒸气质量分数呈弱线性相关，而后期则呈现强线性关系。不凝性气体含量对气泡形状演化和上升速度有显著影响。具体来说，凝结最初集中在气泡的侧面，然后在后期逐渐向顶部和底部转移。在过冷度和蒸气质量分数较高的条件下，气泡内部流动表现出明显的无序特征。

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引用次数: 0

Diagnosis of water hammer conditions in multi-pipeline system using data-driven and genetic algorithm approach 基于数据驱动和遗传算法的多管道系统水锤状态诊断

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.pnucene.2026.106274

Qingyin Zeng , Cheng Peng , Jiang Wu , Jian Deng

While existing experimental investigations and numerical simulations have preliminarily elucidated the multi-factor coupled mechanism governing Condensation-Induced-Water-Hammer (CIWH), the conventional threshold-based alarm approach predominantly relies on single-parameter criteria, e.g. typically pressure, which fails to effectively capture the early warning signs of CIWH. Concurrently, the genetic algorithm, recognized for its applicability in complex systems, remains relatively underutilized in the rapid identification of hazardous operating conditions, resulting in a technical gap where mechanistic understanding is disconnected from practical prevention and control requirements. This study investigates CIWH in the feedwater pipelines of the secondary circuit deaerator in nuclear power plants, where subcooled feedwater mixes with steam. A numerical simulation method based on the NUMAP code is developed to capture the underlying dynamics. Pilot study reveals that feedwater flow rate and pipe diameter jointly regulate the balance between inertial and frictional forces. This interaction shapes flow distribution and void fraction, which in turn influence the intensity of CIWH. Building upon this insight, a systematic analysis is conducted to quantify the effect of feedwater flow rate and pipe diameter on pressure variation rates, with peak values occurring at approximately 84.5 kg/s and 506 mm, respectively. To further identify hazardous operating conditions, a genetic algorithm (GA) is employed with different fitness functions. The results demonstrate that the relative change rate outperforms other metrics, whereas the mean absolute change and standard deviation show certain deviations, and the coefficient of variation is the least effective. This study confirms the effectiveness of the genetic algorithm in identifying hazardous operating conditions of CIWH under complex coupled scenarios and provides a feasible approach for predictive safety control and operational risk assessment in nuclear power plants.

虽然现有的实验研究和数值模拟已经初步阐明了冷凝诱发水锤（CIWH）的多因素耦合机理，但传统的基于阈值的报警方法主要依赖于单参数标准，如典型的压力，不能有效地捕捉到冷凝诱发水锤（CIWH）的预警信号。同时，遗传算法在复杂系统中的适用性得到认可，但在快速识别危险操作条件方面仍未得到充分利用，导致机械理解与实际预防和控制要求脱节的技术差距。本文研究了核电站二回路除氧器给水管道中过冷给水与蒸汽混合的CIWH。提出了一种基于NUMAP代码的数值模拟方法来捕捉其底层动态。初步研究表明，给水量和管径共同调节惯性力和摩擦力的平衡。这种相互作用决定了流动分布和空隙率，进而影响CIWH的强度。在此基础上，进行了系统分析，量化了给水流量和管径对压力变化率的影响，峰值分别约为84.5 kg/s和506 mm。为了进一步识别危险工况，采用了不同适应度函数的遗传算法（GA）。结果表明，相对变化率优于其他指标，而平均绝对变化和标准差存在一定的偏差，变异系数效果最差。本研究证实了遗传算法在复杂耦合情景下识别CIWH危险工况的有效性，为核电厂安全预测控制和运行风险评估提供了一种可行的方法。

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引用次数: 0

Prediction of post-dryout heat transfer based on physics-embedded machine learning with Bayesian optimization algorithm 基于嵌入物理的机器学习与贝叶斯优化算法的干燥后传热预测

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.pnucene.2026.106263

Meiqi Song , Zuokai Chen , Jianhua Xia , Haozhe Li , Wei Xu , Xiaojing Liu

In nuclear power system, encountering the post-dryout heat transfer region can lead to severe heat transfer deterioration. Therefore, it is of great importance to give accurate prediction to post-dryout heat transfer. This study developed a new Physics-Embedded Machine Learning (PEML) framework to predict post-dryout heat transfer, addressing the limitations of traditional "black box" models by integrating physical constraints. Thirteen independent dimensionless parameters (e.g.,

R e_{TP}

,

\Pr_{w}

), i.e., input features, and Nusselt number

N u_{c}

are derived to present the physical heat transfer mechanism. The

N u_{c}

is embedded into the loss function in proportionality or subtractive relationship, i.e., PEML(Nu_exp/Nu_c) and PEML(Nu_exp-Nu_c). The prediction capability of PEML models are better than traditional correlations. The PEML(Nu_exp/Nu_c) model achieves the best prediction capability with the mean error of 0.0005 and RMS error of 0.007 on the testing dataset from Becker's PDO experiments. It is indicated that increasing the number of input features generally improved model performance, especially the generalizability. The PEML framework successfully embeds heat transfer physics, bridging data-driven models and physical insights, offering a robust prediction tool for heat transfer.

在核电系统中，遇到干后换热区会导致严重的换热恶化。因此，对干燥后传热进行准确预测具有重要意义。本研究开发了一种新的物理嵌入式机器学习（PEML）框架来预测干燥后的传热，通过集成物理约束来解决传统“黑箱”模型的局限性。导出了13个独立的无量纲参数（如ReTP， Prw），即输入特征和努塞尔数Nuc来表示物理传热机理。Nuc按比例或相减关系嵌入到损失函数中，即PEML（Nuexp/Nuc）和PEML（Nuexp-Nuc）。PEML模型的预测能力优于传统的相关模型。在Becker的PDO实验数据集上，PEML（Nuexp/Nuc）模型的预测能力最好，平均误差为0.0005，均方根误差为0.007。结果表明，增加输入特征的数量通常会提高模型的性能，尤其是泛化能力。PEML框架成功嵌入了传热物理，桥接了数据驱动模型和物理见解，为传热提供了强大的预测工具。

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引用次数: 0

Assessment of interfacial area concentration models in RELAP5 and TRACE in application to bubbly and cap-bubbly flows in a large square channel 基于RELAP5和TRACE的界面面积浓度模型在大方形通道气泡流和帽状气泡流中的应用评价

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-23 DOI: 10.1016/j.pnucene.2026.106272

Haomin Sun , Takashi Hibiki

A thermal-hydraulic system analysis code is essential for evaluating the performance and safety of nuclear reactors. RELAP5 and TRACE are two widely used system codes that employ the two-fluid model to simulate gas-liquid two-phase flows. Interfacial area concentration (IAC) is one of the key flow parameters required to close the two-fluid model, and its modeling performance directly affects the simulation accuracy. Given its importance, several IAC models have been developed using existing experimental databases for various flow channels, including pipes, rod bundles, and rectangular channels. However, no IAC model development has been conducted on large square channels, despite their significance in the design of advanced light-water nuclear reactors, such as the ESBWR. To address the need for reliably analyzing two-phase flows in large square channels, the two-group (2G) IAC models in RELAP5 and TRACE were evaluated using a large square channel experiment that included 2G flow measurements for bubbly and cap-bubbly flows. Here, the 2G approach refers to a method that classifies bubbles into two bubble groups based on the drag coefficient acting on bubbles. Significant prediction errors were identified in both sub-models that comprise the RELAP5 and TRACE IAC models: the 2G void fraction (VF) model and the VF-to-IAC model (calculating IAC from VF). A recently developed 2G drift-flux correlation was recommended to improve the prediction accuracy of the 2G VF. The VF-to-IAC models in RELAP5 and TRACE were modified based on the experimental data. The mean relative deviations of IAC prediction for large square channels were −4 % and −1 % using the modified RELAP5 and TRACE IAC models, respectively, and improved from 73 % and −26 % using their respective default models.

热液系统分析规范是评价核反应堆性能和安全性的必要工具。RELAP5和TRACE是两种广泛使用的系统代码，采用双流体模型来模拟气液两相流动。界面面积浓度（IAC）是关闭双流体模型所需的关键流动参数之一，其建模性能直接影响仿真精度。鉴于其重要性，利用现有的实验数据库开发了几种IAC模型，用于各种流动通道，包括管道、杆束和矩形通道。然而，尽管大型方形通道在先进轻水核反应堆（如ESBWR）的设计中具有重要意义，但尚未对大型方形通道进行IAC模型开发。为了解决可靠地分析大方形通道中两相流的需求，RELAP5和TRACE中的两组（2G） IAC模型使用大方形通道实验进行了评估，其中包括气泡和帽状气泡流动的2G流量测量。这里的2G法是指根据作用在气泡上的阻力系数将气泡分为两组的方法。在包括RELAP5和TRACE IAC模型的两个子模型中发现了显著的预测误差：2G空隙率（VF）模型和VF-to-IAC模型（从VF计算IAC）。为了提高2G VF的预报精度，建议采用最近发展的2G漂移通量相关方法。根据实验数据对RELAP5和TRACE中的VF-to-IAC模型进行了修正。使用改进的RELAP5和TRACE IAC模型，大方形通道IAC预测的平均相对偏差分别为- 4%和- 1%，而使用各自的默认模型，IAC预测的平均相对偏差分别为73%和- 26%。

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引用次数: 0

Comparative analysis of the ACME and APEX thermal-hydraulic test facilities for advanced passive PWRs ACME与APEX先进被动压水堆热水力试验装置的对比分析

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.pnucene.2026.106266

Chengcheng Deng, Chengzhi Li, Junxiao Yang, Zongyang Li

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.

为了对先进核反应堆进行安全评价，有必要设计和建造规模热水力试验设施，以复制原型反应堆的关键现象和过程。先进堆芯冷却机理实验（ACME）和先进电站实验（APEX）是为先进无源堆设计和建造的两个典型的一体化热工试验设备。APEX测试设施是为AP1000建造的，而ACME测试设施是为CAP1400在中国建造的，并被选为国际标准问题51号（ISP-51）项目的基准设施。本研究从定性和定量两个角度对ACME和APEX试验设备进行了比较分析。一方面，结合实验数据和ACME和APEX的仿真结果，对比分析了SBLOCA瞬态过程中关键参数的变化曲线；另一方面，采用系统级尺度分析方法，定量比较了SBLOCA暂态过程中不同阶段关键现象的无因次数。结果表明，ACME和APEX都是测试被动安全系统在SBLOCA瞬态过程下行为的良好规模的热水力试验设施。此外，ACME和APEX试验装置在SBLOCA瞬态过程中后期表现出良好的相似性。通过ACME和APEX设备的对比分析，本研究可为先进核反应堆整体热水力试验设备的标度设计和实验数据的交互验证提供指导。

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引用次数: 0

RELAP5-3D simulation of natural circulation start-up and station blackout benchmark for a NuScale-like iPWR 类似于nuscale的iPWR自然循环启动和站点停电基准的RELAP5-3D模拟

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.pnucene.2026.106265

Vincenzo Zingales , Francesco D'Auria , Yassin A. Hassan

This paper discusses the key features of integral Pressurized Water Reactors (iPWRs) including Helical Coil Steam Generators (HCSGs) and associated correlations. The objective is to highlight design challenges and system level modeling strategies.

A RELAP5-3D ver. 4.4.2 nodalization of a NuScale-like iPWR was created based on publicly available literature and, where necessary, design assumptions. The model was qualified under steady-state conditions against the NuScale Final Safety Analysis Report approved as part of the Design Certification Application (DCA) in 2020.

Because HCSGs are prone to instabilities, a start-up procedure was simulated to test the model response across a range of operating parameters. Primary flow and temperature results have been found to be consistent with DCA data at reduced power. However, at power levels below 60%, Type-II Density Wave Oscillations (DWOs) occurred in the HCSGs tubes. Analysis of subcooling and phase change numbers informed of potential mitigation strategies; however, stable performance at low power could not be obtained together with steam superheat and constant primary average temperature.

The main outcome of the present paper is the presentation of a collaborative benchmark internal to the Texas A&M University in which a station blackout scenario for a NuScale-like iPWR was simulated. Results obtained from RELAP5-3D were compared with those from TRACE, the NuScale Simulator, and DCA data, demonstrating strong qualitative agreement but highlighting quantitative discrepancies primarily due to geometric and correlation differences among the different models. RELAP5-3D simulations have specifically highlighted the occurrence of Type-I DWO if the ECCS is not timely activated.

本文讨论了包括螺旋盘管蒸汽发生器（hcsg）在内的整体式压水堆（iPWRs）及其相关特性。目标是强调设计挑战和系统级建模策略。RELAP5-3D版本。4.4.2基于公开可用的文献和必要时的设计假设，创建了类似于nuscale的iPWR的nodalization。该模型在稳态条件下符合NuScale最终安全分析报告，该报告是2020年设计认证申请（DCA）的一部分。由于hcsg容易不稳定，因此模拟了启动过程，以测试模型在一系列操作参数下的响应。一次流量和温度的结果已经发现与DCA数据在降低功率一致。然而，当功率低于60%时，hcsg管中出现了ii型密度波振荡（dwo）。了解潜在缓解战略的过冷和相变数分析；然而，在蒸汽过热度和一次平均温度不变的情况下，低功率下无法获得稳定的性能。本论文的主要成果是介绍了德克萨斯农工大学内部的协作基准，其中模拟了nuscal类iPWR的站点停电场景。从RELAP5-3D获得的结果与TRACE、NuScale Simulator和DCA数据进行了比较，结果表明定性一致，但突出了定量差异，主要是由于不同模型之间的几何和相关性差异。RELAP5-3D模拟特别强调了如果ECCS没有及时激活，就会发生i型DWO。

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引用次数: 0

Study on hydrodynamic characteristics of symmetric two-droplet impact on film using lattice Boltzmann method 用晶格玻尔兹曼方法研究对称双液滴撞击薄膜的水动力特性

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-20 DOI: 10.1016/j.pnucene.2026.106269

Huifang Zhang, Jian Yu, Yapei Zhang, Shihao Wu, Wenxi Tian, Suizheng Qiu, Guanghui Su

Droplet impact on a liquid film is ubiquitous omnipresent and very fundamental in nature and industrial. For instance, in the spray cooling of the lower head of reactor pressure vessels, the method can enhance the safety margin of reactors. Extensive research has been carried out on the vertical impact of multiple droplets or single droplet on liquid films. However, the dynamical characteristics of multiple droplets impacting inclined liquid films remain insufficiently understood. Moreover, simulation approaches have predominantly concentrated on the Volume of Fluid (VOF) method. Therefore, this study attempts to conduct an in-depth numerical investigation of this phenomenon using the lattice Boltzmann method (LBM). A computational model was developed based on the Q3D27 and validated through benchmark cases involving single-droplet impacts on liquid films under both vertical and oblique conditions. The model accurately predicted key characteristics such as the outer diameter of the crown splash and the upstream crown radius. Based on the validated model, simulations of oblique impacts by dual droplets on a thin liquid film were conducted. The interfacial evolution was systematically analyzed, including the formation and development of crown splashes as well as the dynamics of intermediate thin-film jets. Furthermore, the Plateau-Rayleigh instability theory was employed to investigate the breakup mechanisms of liquid columns under varying impact angles and velocities. The fluid dynamic interactions between the two droplets under oblique impact conditions were also examined in detail, revealing complex flow behaviors relevant to multiphase flow dynamics.

液滴对液膜的影响在自然界和工业中是无处不在的，也是非常重要的。例如，在反应堆压力容器下封头喷雾冷却中，该方法可以提高反应堆的安全裕度。人们对多液滴或单液滴对液膜的垂直影响进行了广泛的研究。然而，多液滴撞击倾斜液膜的动力学特性仍未得到充分的了解。此外，模拟方法主要集中在流体体积法（VOF）上。因此，本研究试图利用晶格玻尔兹曼方法（LBM）对这一现象进行深入的数值研究。基于Q3D27建立了计算模型，并通过垂直和倾斜条件下单液滴撞击液膜的基准案例进行了验证。该模型准确地预测了关键特性，如冠飞溅外径和上游冠半径。在验证模型的基础上，对双液滴在薄膜上的斜碰撞进行了模拟。系统地分析了界面演化过程，包括冠状飞溅的形成和发展以及中间薄膜射流的动力学过程。利用高原-瑞利不稳定性理论研究了不同冲击角度和冲击速度下液柱破碎机理。本文还详细研究了两液滴在斜碰撞条件下的流体动力学相互作用，揭示了与多相流动力学相关的复杂流动行为。

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引用次数: 0

Effect of channel geometry on the onset of significant void and void fraction profiles of vertical upward boiling flows 通道几何形状对垂直向上沸腾流动中显著空隙和空隙分数分布的影响

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-19 DOI: 10.1016/j.pnucene.2026.106256

Shichang Dong, Takashi Hibiki

The assessment of heat transfer performance characteristics relies on the accurate prediction of void fraction in heat transfer systems. Existing typical void fraction models rely on the determination of the point of onset of significant void (OSV), which limits their ability to estimate the development of void fraction before the OSV point. To address this limitation, the present study first developed a flow quality model for the subcooled boiling region. This model was then extended to the saturated boiling region, resulting in a flow quality model applicable across the entire boiling region. Subsequently, this model was integrated with a drift-flux correlation (DFC) to formulate a predictive model for axial void fraction profiles in vertical boiling flows in annular channels. The validation of the developed model was conducted using a comprehensive experimental database including 1300 data points from five independent sources with water as the working fluid. Results showed that, compared to the existing typical model, the developed model could successfully predict the evolution of the axial void fraction upstream of the OSV point. This model also demonstrated superior accuracy in predicting the OSV point and downstream void fraction behavior, thereby significantly enhancing the prediction accuracy of the axial void fraction profiles. Furthermore, it accurately predicted the influence of key thermal-hydraulic parameters on the void fraction profiles. This paper also revealed the impact of channel geometry by comparing OSV and void fraction profiles across circular, rectangular, and annular channels.

传热性能特性的评估依赖于传热系统中空隙率的准确预测。现有的典型孔隙分数模型依赖于有效孔隙（OSV）起始点的确定，这限制了它们对OSV点之前孔隙分数发展的估计能力。为了解决这一限制，本研究首先建立了过冷沸腾区的流动质量模型。然后将该模型推广到饱和沸腾区域，得到了适用于整个沸腾区域的流动质量模型。随后，将该模型与漂移通量相关（DFC）相结合，建立了环形通道垂直沸腾流轴向空隙率分布的预测模型。利用一个综合的实验数据库，包括来自五个独立来源的1300个数据点，以水为工作流体，对所开发的模型进行了验证。结果表明，与现有的典型模型相比，所建立的模型能够较好地预测OSV点上游轴向空隙率的演化。该模型在预测OSV点和下游含气分数行为方面也具有较好的准确性，从而显著提高了轴向含气分数剖面的预测精度。同时，准确预测了关键热液参数对孔隙率分布的影响。本文还通过比较圆形、矩形和环形通道上的OSV和空隙率分布，揭示了通道几何形状的影响。

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引用次数: 0

Numerical study on heat transfer enhancement of LBE flow in semicircular-fin fuel bundles 半圆形翅片燃料束中LBE流动强化传热的数值研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-19 DOI: 10.1016/j.pnucene.2026.106253

Qian Li , Siwei Cai , Shengcai Zhang , Xuechen Liu , Nianmei Zhang , Chen Hu , Xian Zeng , Jianchuang Sun , Weihua Cai

In this study, numerical simulation methods were employed to investigate the flow and heat transfer characteristics of liquid lead-bismuth eutectic (LBE) alloy in novel fuel assemblies with different fin winding directions of the fuel rod bundle. The axial and circumferential thermal-hydraulic parameter data of subchannels and fuel rods were extracted through micro-segment methods and sub-channel partitioning. The research results showed that changing the fin winding direction had a minor effect on the pressure drop of the fuel assembly, while the heat transfer coefficient increased by 15 % compared to the original fuel assemblies, demonstrating that the new fuel assemblies effectively enhance the heat transfer capacity of LBE in semicircular-fin rod bundles. Further investigation into the mechanisms of enhanced heat transfer in novel fuel assemblies reveals that altering the orientation of the fins results in co-directional flow of the LBE. This co-directional flow can increase the secondary flow velocity while maintaining the Q invariant as a positive value, thereby reducing the occurrence of rotational flow and enhancing heat transfer capabilities. This study offers new research perspectives on the design of lead-bismuth fast reactor fuel assemblies and the analysis of their thermal-hydraulic characteristics, which are of significant importance for the structural design of lead-bismuth fast reactors and for the mechanistic research of nuclear reactors.

本文采用数值模拟方法研究了液态铅铋共晶合金在不同燃料棒束翅片缠绕方向的新型燃料组件中的流动和传热特性。通过微段法和子通道划分提取子通道和燃料棒的轴向和周向热工参数数据。研究结果表明，改变翅片缠绕方向对燃料组件的压降影响较小，但换热系数较原燃料组件提高了15%，表明新型燃料组件有效提高了半圆翅片棒束内LBE的换热能力。对新型燃料组件中强化传热机制的进一步研究表明，改变翅片的方向会导致LBE的共向流动。这种共向流动可以在保持Q不变量为正值的同时增加二次流速度，从而减少旋转流动的发生，增强换热能力。本研究为铅铋快堆燃料组件的设计及热水力特性分析提供了新的研究视角，对铅铋快堆的结构设计和核反应堆的机理研究具有重要意义。

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引用次数: 0