Progress in Nuclear Energy最新文献_第3页

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%。

{"title":"Assessment of interfacial area concentration models in RELAP5 and TRACE in application to bubbly and cap-bubbly flows in a large square channel","authors":"Haomin Sun , Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106272","DOIUrl":"10.1016/j.pnucene.2026.106272","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106272"},"PeriodicalIF":3.2,"publicationDate":"2026-01-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039147","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}

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。

{"title":"Comparative analysis of the ACME and APEX thermal-hydraulic test facilities for advanced passive PWRs","authors":"Chengcheng Deng, Chengzhi Li, Junxiao Yang, 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":"2026-01-21","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}

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。

{"title":"RELAP5-3D simulation of natural circulation start-up and station blackout benchmark for a NuScale-like iPWR","authors":"Vincenzo Zingales , Francesco D'Auria , Yassin A. Hassan","doi":"10.1016/j.pnucene.2026.106265","DOIUrl":"10.1016/j.pnucene.2026.106265","url":null,"abstract":"<div><div>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.</div><div>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.</div><div>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.</div><div>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.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106265"},"PeriodicalIF":3.2,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039089","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}

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。

{"title":"Study on hydrodynamic characteristics of symmetric two-droplet impact on film using lattice Boltzmann method","authors":"Huifang Zhang, Jian Yu, Yapei Zhang, Shihao Wu, Wenxi Tian, Suizheng Qiu, Guanghui Su","doi":"10.1016/j.pnucene.2026.106269","DOIUrl":"10.1016/j.pnucene.2026.106269","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106269"},"PeriodicalIF":3.2,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039088","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}

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。

{"title":"Effect of channel geometry on the onset of significant void and void fraction profiles of vertical upward boiling flows","authors":"Shichang Dong, Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106256","DOIUrl":"10.1016/j.pnucene.2026.106256","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106256"},"PeriodicalIF":3.2,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039087","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}

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。

{"title":"Numerical study on heat transfer enhancement of LBE flow in semicircular-fin fuel bundles","authors":"Qian Li , Siwei Cai , Shengcai Zhang , Xuechen Liu , Nianmei Zhang , Chen Hu , Xian Zeng , Jianchuang Sun , Weihua Cai","doi":"10.1016/j.pnucene.2026.106253","DOIUrl":"10.1016/j.pnucene.2026.106253","url":null,"abstract":"<div><div>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 <em>Q</em> 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.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106253"},"PeriodicalIF":3.2,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039148","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}

引用次数: 0

Modeling flow quality for predicting axial void fraction development of boiling flows in rectangular channels at elevated pressure conditions 高压条件下矩形通道沸腾流轴向空隙率发展的流动质量模拟

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.2026.106255

Shichang Dong, Takashi Hibiki

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.

在矩形通道内沸腾流动的传热传质分析中，空隙率的准确估计起着关键作用。现有经典孔隙分数模型采用有效空隙起始点（OSV）作为轴向孔隙分数发育的起始位置，这限制了其准确预测有效空隙起始点周围孔隙分数的能力。为了解决这一限制，首先开发了过冷沸腾流的流动质量相关性，然后扩展到覆盖整个沸腾状态。在此基础上，结合全范围漂移-通量相关，建立了矩形通道内沸腾流的轴向空隙率模型。该模型使用由八个独立来源编制的广泛实验数据库进行了严格验证。结果表明，与经典模型相比，该模型能更准确地预测OSV点上游和下游的轴向空隙率演化。此外，相对于Saha-Zuber相关，它在预测OSV点热平衡质量方面的能力有所提高。此外，该模型准确地反映了几个关键参数对空隙率分布的影响，表明该模型适用于矩形通道内的沸腾流动分析。

{"title":"Modeling flow quality for predicting axial void fraction development of boiling flows in rectangular channels at elevated pressure conditions","authors":"Shichang Dong, 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":"2026-01-19","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}

引用次数: 0

Experimental study on geyser boiling parameter characteristics in vertical sodium heat pipes 垂直钠热管间歇泉沸腾参数特性实验研究

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.2026.106270

Youlan Yuan , Zaiyong Ma , Yugao Ma , Zheng Zhou , Luteng Zhang , Simiao Tang , Liangming Pan

State of the liquid pool is vital to the stable operation of vertical sodium heat pipes. Geyser boiling in the pool region of vertical sodium heat pipes would induce significant temperature and pressure oscillations, which could negatively impact heat transfer and structural safety. In this paper, experiment was conducted to investigate the effects of input power, filling ratio, and evaporator length on the characteristics of geyser boiling in a vertical sodium heat pipe, including average oscillation period, amplitude, bubble size, and vapor generation amount. The results indicated that both the average oscillation period and amplitude decreased with increasing heating power, as the system at lower heating powers tended to exhibit non-periodic geyser boiling, leading to larger oscillation periods and amplitudes. A higher filling ratio could significantly shorten the oscillation period, indicating that more oscillation cycles occur within the same time interval. Theoretical analysis combined with experimental observation revealed that the bubble diameter during geyser boiling mainly ranged from 0.9 to 2.3 D (inner diameter of the heat pipe). Increasing filling ratio and evaporator length inhibited the formation of large bubbles, resulting in significantly reduced system fluctuations. Likewise, when the filling ratio was increased from 41.67 % to 83.33 %, the variation in bubble size over time became smaller. A dimensionless vapor generation factor f, based on vapor production inferred from temperature oscillations, was proposed to characterize vapor generation amount of each bubble release. The factor f remained generally low and exhibited no simple monotonic relationship with heating power. It showed enhanced instability at lower filling ratios and a non-monotonic dependence on evaporator length, reaching a peak value near 350 mm.

液池的状态对垂直钠热管的稳定运行至关重要。垂直钠热管池区间歇泉沸腾会引起明显的温度和压力振荡，对传热和结构安全产生不利影响。本文通过实验研究了输入功率、填充比和蒸发器长度对垂直钠热管中间歇泉沸腾特性的影响，包括平均振荡周期、振幅、气泡大小和蒸汽生发量。结果表明，随着加热功率的增加，平均振荡周期和振幅均减小，较低加热功率时系统出现非周期性间歇泉沸腾，振荡周期和振幅较大；充填率越高，振荡周期越短，表明在相同的时间间隔内振荡周期越多。理论分析与实验观察相结合表明，间歇泉沸腾过程中的气泡直径主要在0.9 ~ 2.3 D（热管内径）之间。增加填充比和蒸发器长度可以抑制大气泡的形成，从而显著降低系统波动。同样，当填充率从41.67%增加到83.33%时，气泡尺寸随时间的变化变小。提出了一个无量纲的蒸汽产生因子f，该因子基于温度振荡推断的蒸汽产生，来表征每次气泡释放的蒸汽产生量。因子f总体保持较低，与加热功率之间不存在简单的单调关系。在较低的填充比下，它表现出增强的不稳定性，并且与蒸发器长度非单调依赖，在350 mm附近达到峰值。

{"title":"Experimental study on geyser boiling parameter characteristics in vertical sodium heat pipes","authors":"Youlan Yuan , Zaiyong Ma , Yugao Ma , Zheng Zhou , Luteng Zhang , Simiao Tang , Liangming Pan","doi":"10.1016/j.pnucene.2026.106270","DOIUrl":"10.1016/j.pnucene.2026.106270","url":null,"abstract":"<div><div>State of the liquid pool is vital to the stable operation of vertical sodium heat pipes. Geyser boiling in the pool region of vertical sodium heat pipes would induce significant temperature and pressure oscillations, which could negatively impact heat transfer and structural safety. In this paper, experiment was conducted to investigate the effects of input power, filling ratio, and evaporator length on the characteristics of geyser boiling in a vertical sodium heat pipe, including average oscillation period, amplitude, bubble size, and vapor generation amount. The results indicated that both the average oscillation period and amplitude decreased with increasing heating power, as the system at lower heating powers tended to exhibit non-periodic geyser boiling, leading to larger oscillation periods and amplitudes. A higher filling ratio could significantly shorten the oscillation period, indicating that more oscillation cycles occur within the same time interval. Theoretical analysis combined with experimental observation revealed that the bubble diameter during geyser boiling mainly ranged from 0.9 to 2.3 D (inner diameter of the heat pipe). Increasing filling ratio and evaporator length inhibited the formation of large bubbles, resulting in significantly reduced system fluctuations. Likewise, when the filling ratio was increased from 41.67 % to 83.33 %, the variation in bubble size over time became smaller. A dimensionless vapor generation factor <em>f</em>, based on vapor production inferred from temperature oscillations, was proposed to characterize vapor generation amount of each bubble release. The factor <em>f</em> remained generally low and exhibited no simple monotonic relationship with heating power. It showed enhanced instability at lower filling ratios and a non-monotonic dependence on evaporator length, reaching a peak value near 350 mm.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106270"},"PeriodicalIF":3.2,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039086","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}

引用次数: 0

Analyzing the sources of radioactive waste from Small Modular Reactors (SMR) 小型模块化反应堆（SMR）放射性废物来源分析

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

Progress in Nuclear Energy

Pub Date : 2026-01-17 DOI: 10.1016/j.pnucene.2026.106239

Mumtaz Khan , Hu Penghua , Jie Niu , Xu Lechang , Wang Yalan , Thaqal M. Alhuzaymi

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.

本综述使用与储存库相关的指标（1/10/100年的封装规模衰变热、可移动裂变产物库存/形态、可裂变物含量/临界边际和二次废物量），对主要SMR系列——ipwr（如NuScale）、PHWR/ candu级、HTGR （x -100）、SFR（如ARC/ natium级）和MSR（如IMSR）——与大型压水堆基线进行了废物产生和后端挑战的批判性比较。定量地，独立评估表明，由于更高的中子泄漏和反应性冷却剂，与千兆瓦级轻水堆相比，几个小堆概念可以使每兆瓦时所需管理的总废物增加~ 2-30倍；具有代表性的iPWR每单位电的SNF排放量比大型压水堆多1.7倍，并有来自紧凑整体容器的额外活性金属ILW。相比之下，Xe-100实现了非常高的燃耗，SNF质量降低了~ 75%，100年衰变热降低了35 - 50%，但代价是由GTCC石墨主导的SNF几何体积增加了~ 10-12倍。SFRs可能会减少每兆瓦时SNF质量，但会产生富含truu的钠键合燃料，这种燃料通常需要热处理，并产生明显的钠清除残留物。candu级smr降低了长寿命TRU库存，但增加了SF封装数量（低燃耗），并且需要一流的氚系统。MSRs/IMSR将高废渣转化为含燃料盐和废气介质，使盐匹配的废物形式资格（例如玻璃粘合陶瓷）具有决定性。我们综合了设计特定的废物驱动因素，确定了数据缺口（例如，FOAK→NOAK，每兆瓦时基准），并概述了研究方向：(i)泄漏/活化减少，（ii）石墨和氚质量平衡管理，（iii）钠中和和er-盐废物形式，以及（iv）盐特定固定路线的确定。总的来说，证据倾向于根据面向储存库的指标来评估smr，而不是大批量索赔，除非后端协同设计证明有效，否则一些概念可能会增加每兆瓦时的废物负担。

{"title":"Analyzing the sources of radioactive waste from Small Modular Reactors (SMR)","authors":"Mumtaz Khan , Hu Penghua , Jie Niu , Xu Lechang , Wang Yalan , 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":"2026-01-17","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}

引用次数: 0

Numerical study on pressure oscillation characteristics of unstable steam jet condensation through round/obround side-hole spargers 圆/外侧孔喷射器不稳定蒸汽射流冷凝压力振荡特性的数值研究

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

Progress in Nuclear Energy

Pub Date : 2026-01-17 DOI: 10.1016/j.pnucene.2025.106202

Jiayu Xiao, Zhongning Sun, Haozhi Bian, Zhaoming Meng

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.

不稳定的蒸汽喷射冷凝会引起强烈的压力振荡。对圆孔和圆孔喷砂器的不稳定蒸汽射流进行了瞬态数值模拟。该模型有效地捕获了两种类型孔的主频率差异。讨论了流体域压力波动特性与气泡演化过程的关系。压力峰值出现在二次气泡破裂位置附近。在相同的热条件下，圆孔产生的时间平均气泡体积是圆孔的1.22倍，时间平均穿透深度是圆孔的0.96倍，凝结振荡周期是圆孔的1.16倍。利用112个压力监测点的数据重构了流体域的压力分布。流体域中存在压力峰区，各位置压力峰的大小与距二次气泡破裂位置的距离几乎成反比。喷淋表面最大压力峰出现在距孔边0.5 ~ 1.5倍直径处。随着蒸汽质量通量的增大和水过冷度的增大，区域内最大峰值压力增大，同时对喷淋壁的压力冲击也增大。

{"title":"Numerical study on pressure oscillation characteristics of unstable steam jet condensation through round/obround side-hole spargers","authors":"Jiayu Xiao, Zhongning Sun, Haozhi Bian, 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":"2026-01-17","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}

引用次数: 0