Progress in Nuclear Energy最新文献_第4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 : 2026-04-01 Epub 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

Experimental and numerical study of particle resuspension on turbine surface with film cooling 气膜冷却涡轮表面颗粒再悬浮的实验与数值研究

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

Progress in Nuclear Energy

Pub Date : 2026-04-01 Epub Date: 2026-02-03 DOI: 10.1016/j.pnucene.2026.106285

Xiaozhong Wang , Qi Sun , Wei Peng , Yinhai Zhu , Peixue Jiang , Suyuan Yu

Resuspension significantly influences the transport and deposition of graphite particles within the turbine of high-temperature gas-cooled reactor (HTGR). This work involved setting up a visualized experimental setup to observe particle resuspension under film cooling conditions. High-speed cameras and hot-wire anemometers were employed to measure the dynamic process of particle resuspension and the characteristics of the near-wall flow field. Based on the torque balance model, a particle resuspension model suitable for the HTGR turbine conditions was developed, and the accuracy of the model was verified with experimental results. Combining experimental measurements and numerical simulations, the mechanism by which film cooling affects particle resuspension was investigated, and the dominant role of counter-rotating vortices was revealed. Furthermore, the influence of key factors such as flow velocity, blowing ratio, and particle size on resuspension characteristics was systematically analyzed. Finally, the particle deposition process and resuspension process were coupled for analysis, and the distribution characteristics of particle deposition on the turbine surface under film cooling conditions were studied. The findings from this study not only offer experimental insights into particle resuspension behavior but also contribute valuable guidance for the design and safety assessment of HTGR helium turbines.

重悬浮对高温气冷堆（HTGR）涡轮内石墨颗粒的迁移和沉积有重要影响。这项工作包括建立一个可视化的实验装置来观察膜冷却条件下的颗粒再悬浮。采用高速摄像机和热线风速仪测量了颗粒重悬浮的动态过程和近壁流场特征。在转矩平衡模型的基础上，建立了适用于HTGR涡轮工况的颗粒重悬浮模型，并用实验结果验证了模型的准确性。结合实验测量和数值模拟，研究了气膜冷却对颗粒重悬浮的影响机理，揭示了对旋涡的主导作用。系统分析了流速、吹气比、粒度等关键因素对再悬浮特性的影响。最后，对颗粒沉积过程和重悬浮过程进行耦合分析，研究了气膜冷却条件下涡轮表面颗粒沉积的分布特征。该研究结果不仅为粒子再悬浮行为提供了实验见解，而且为HTGR氦涡轮的设计和安全性评估提供了有价值的指导。

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引用次数: 0

Modeling terminal velocity and aspect ratio of a single bubble in distorted-particle and cap bubble regimes for pool scrubbing simulation 对池擦洗模拟中扭曲粒子和帽泡状态下单个气泡的终端速度和长径比进行建模

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

Progress in Nuclear Energy

Pub Date : 2026-04-01 Epub Date: 2026-02-05 DOI: 10.1016/j.pnucene.2026.106267

Yuria Okagaki , Takashi Hibiki

Pool scrubbing is an essential filtration process that prevents the release of radioactive aerosols into the environment during severe accidents at nuclear reactors. During such events, a mixture of steam and/or various noncondensable gases containing radioactive aerosols is injected into the pool, where it forms bubbles. During aerosol particle transport, bubble dynamics play a notable role in determining the efficiency of particle removal within the bubbles. Clarifying these mechanisms is crucial for accurately modeling pool scrubbing processes. In practice, pool scrubbing codes based on the lumped parameter (LP) approach incorporate particle removal models; however, key hydrodynamic parameters—such as bubble rise (terminal) velocity and bubble aspect ratio—are modeled using simplified assumptions. The present study aimed to verify the validity of existing correlations for terminal velocity and bubble aspect ratio, with the long-term objective of proposing improved models. As part of this fundamental research, the behavior of a single bubble rising in a stagnant liquid, simulating a pool environment, was visualized through computational fluid dynamics (CFD) simulations employing an interface tracking/capturing method. The effective aspect ratio was calculated based on the bubble interfacial area and volume. Furthermore, correlations for single-bubble terminal velocity and aspect ratio were developed, taking into account the transition in bubble shape as the volume-equivalent diameter increased. The findings from the present study are expected to provide a foundation for the development and evaluation of future standalone pool scrubbing codes, including improved correlations for terminal velocity for bubble swarm, bubble aspect ratio, and other relevant hydrodynamic parameters.

在核反应堆发生严重事故时，池擦洗是一个重要的过滤过程，可以防止放射性气溶胶释放到环境中。在这种情况下，蒸汽和/或含有放射性气溶胶的各种不可冷凝气体的混合物被注入池中，在那里形成气泡。在气溶胶粒子传输过程中，气泡动力学对粒子在气泡内的去除效率起着重要的作用。澄清这些机制对于准确模拟池擦洗过程至关重要。在实践中，基于集总参数（LP）方法的池擦洗代码包含颗粒去除模型；然而，关键的水动力参数-如气泡上升（末端）速度和气泡长径比-使用简化的假设进行建模。本研究旨在验证终端速度和气泡长径比的现有相关性的有效性，并提出改进的长期目标模型。作为这项基础研究的一部分，通过采用界面跟踪/捕获方法的计算流体动力学（CFD）模拟，模拟了水池环境中单个气泡在停滞液体中上升的行为。根据气泡界面面积和体积计算有效宽高比。此外，考虑到气泡形状随着体积当量直径的增加而发生的转变，建立了单泡末端速度与长径比的相关关系。本研究的结果有望为未来独立池洗涤代码的开发和评估提供基础，包括改进泡群末端速度、泡长径比和其他相关水动力参数的相关性。

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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 : 2026-04-01 Epub 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。

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引用次数: 0

Review of self-powered neutron detectors for reactor instrumentation 反应堆仪器用自供电中子探测器综述

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2026-01-09 DOI: 10.1016/j.pnucene.2025.106230

Jack A. Lanza , Lei R. Cao

Self-powered neutron detectors (SPNDs) have been essential instruments for in-core flux monitoring in light-water reactors (LWRs) for over six decades. This paper provides a comprehensive review of SPND principles, materials, and configurations, along with their roles in the third-generation (Gen-III and Gen-III+) reactor designs. The review compares performance characteristics of common emitter materials such as Rh, V, Co, and Ag, and examines emerging candidates, including Cd, Er, and Hf, for prompt response applications. Developments in modeling, compensation algorithms, and signal processing, including analytical, analog, and digital methods such as Kalman filtering, are summarized to highlight improvements in accuracy and response time. Recent advancements have extended to fast-neutron and gamma-sensitive SPNDs, along with opportunities for integration in digital twin and AI-based reactor monitoring frameworks. The paper also identifies research gaps in detector-level integration, materials optimization, and adaptability for small modular and microreactor environments. These findings underscore the continued importance of SPNDs as reliable, radiation-tolerant sensors within evolving nuclear instrumentation and safety systems.

60多年来，自供电中子探测器一直是轻水堆堆芯内通量监测的重要工具。本文全面回顾了SPND的原理、材料和配置，以及它们在第三代（Gen-III和Gen-III+）反应堆设计中的作用。该综述比较了常见发射极材料的性能特征，如Rh、V、Co和Ag，并研究了新兴的候选材料，包括Cd、Er和Hf，用于快速响应应用。在建模、补偿算法和信号处理方面的发展，包括分析、模拟和数字方法，如卡尔曼滤波，总结了精度和响应时间的改进。最近的进展已经扩展到快中子和伽马敏感spnd，以及集成数字孪生和基于人工智能的反应堆监测框架的机会。论文还指出了探测器级集成、材料优化以及小型模块化和微反应堆环境适应性方面的研究差距。这些发现强调了spnd在不断发展的核仪器和安全系统中作为可靠、耐辐射传感器的持续重要性。

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引用次数: 0

Transformation of montmorillonite under thermal-saline conditions in HLW repositories: A systematic review 热盐条件下高沸石储存库中蒙脱石的转化：系统综述

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2025-12-23 DOI: 10.1016/j.pnucene.2025.106201

Cong Liu , Yong-Gui Chen , Rao-Ping Liao , Zhao Sun , Yu-Cheng Li , Sheng-Fei Cao , Wei-Min Ye

The long-term stability of montmorillonite under near-field thermal and saline conditions in a high-level radioactive waste (HLW) repository is a critical concern as potential transformation of montmorillonite may occur under such circumstances. This review aims to summarize current knowledge on the transformation of montmorillonite under thermal and saline conditions relevant to HLW repositories, highlighting key mechanisms and patterns reported in the literature. Laboratory and in-situ tests indicate that montmorillonite can undergo thermo-saline transformations into illite, kaolinite, chlorite, or other minerals, and may also exhibit changes in negative layer charge. The effects of changes in negative layer charge and interlayer cations on the swelling properties of montmorillonite were analyzed. In addition to temperature and salinity, factors such as montmorillonite type, accessory minerals, pressure, reaction time, and solid/liquid ratio influence these transformations of montmorillonite.

高放废物（HLW）储存库中蒙脱土在近场高温和盐水条件下的长期稳定性是一个关键问题，因为在这种情况下可能发生蒙脱土的潜在转化。这篇综述旨在总结目前关于高沸石储存库在高温和盐水条件下蒙脱土转化的知识，重点介绍了文献报道的关键机制和模式。实验室和现场测试表明，蒙脱土可以经历热盐水转化为伊利石、高岭石、绿泥石或其他矿物，也可能表现出负层电荷的变化。分析了层间负电荷和层间阳离子变化对蒙脱土溶胀性能的影响。除温度和盐度外，蒙脱土类型、辅助矿物、压力、反应时间和固液比等因素也会影响蒙脱土的这些转化。

引用次数: 0

How much residual water remains in a canister after vacuum drying for spent nuclear fuel storage? 在真空干燥乏核燃料储存罐后，剩余的水有多少？

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2026-01-03 DOI: 10.1016/j.pnucene.2025.106231

Ji Hwan Lim , Seung-Hwan Yu , Gyung-sun Chae , Kyung-Wook Shin , Nam-Hee Lee

This study presents a rigorous experimental investigation of the direct quantification of residual water remaining inside spent nuclear fuel (SNF) canisters following vacuum drying—a pivotal operation for limiting long-term corrosion, radiolytic gas generation, and structural degradation during dry storage. Using a purpose-built laboratory facility that replicates realistic canister geometries, we examined initial moisture inventories of 10–77.9 g. Drying employed a carefully modulated pressure-reduction protocol, achieving standard-compliant conditions (<3 torr for 30 min), after which the canisters were backfilled to 2 bar with inert gas. Direct moisture measurements used an AquaVolt 2 electrolytic trace analyzer under controlled helium pressurization, returning highly resolved residual-moisture concentrations in ppmv. Across all scenarios, the final internal moisture converged reproducibly to ∼500–600 ppmv (≈0.15 g), indicating a predictable endpoint under standard-compliant drying, with instrument-traceable repeatability across trials. A bounding comparison placed these values 3–4 × below a conservative worst-case assumption (treating the entire residual head as H2O vapor; ∼0.569 g) yet 15–20 × above estimates based on ambient humidity entrainment (∼0.0089 g). Beyond filling a longstanding empirical gap, these measurements have direct engineering and regulatory significance: they supply defendable inputs for hydrogen source-term, corrosion, and pressure-evolution models; enable risk-informed procedure design and QA/QC traceability; and complement current compliance frameworks (e.g., NUREG-2215; ASTM C1553-21) that rely on pressure-based endpoints. Finally, the integrated methodology—controlled vacuum drying, real-time dew-point tracking, and ultra-trace quantitative moisture analysis—provides a practical blueprint for selective, confirmatory direct-moisture verification in SNF dry-storage operations and is transferable to other high-stakes vacuum-dependent industries.

本研究提出了一项严格的实验研究，对真空干燥后乏核燃料（SNF）罐内剩余水的直接量化进行了研究。真空干燥是限制长期腐蚀、放射性溶解气体产生和干燥储存期间结构降解的关键操作。使用一个专门建造的实验室设施，复制真实的罐几何形状，我们检查了10-77.9 g的初始水分库存。干燥采用精心调整的减压方案，达到符合标准的条件（<； 3torr 30分钟），之后用惰性气体回填到2bar。直接水分测量使用AquaVolt 2电解痕量分析仪在受控氦气加压下，返回高度分解的残余水分浓度，单位为ppmv。在所有情况下，最终的内部水分可重复性地收敛到~ 500-600 ppmv(≈0.15 g)，表明在符合标准的干燥条件下可预测的终点，在试验期间具有仪器可追溯的重复性。边界比较使这些值比保守的最坏情况假设（将整个剩余水头视为H2O蒸汽；~ 0.569 g）低3-4倍，但比基于环境湿度夹带（~ 0.0089 g）的估计高15-20倍。除了填补长期的经验空白之外，这些测量还具有直接的工程和监管意义：它们为氢源期、腐蚀和压力演化模型提供了可靠的输入；确保风险知情的程序设计和QA/QC可追溯性；并补充当前依赖压力端点的合规框架（例如，nurg -2215; ASTM C1553-21）。最后，集成的方法-控制真空干燥，实时露点跟踪和超痕量定量水分分析-为SNF干储存操作中的选择性，验证性直接水分验证提供了实用的蓝图，并可转移到其他高风险的真空依赖行业。

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引用次数: 0

Experimentation and simulation research on foreign object wear of cladding-rods interface in high-temperature pressurized water 高温加压水中包层-棒界面异物磨损试验与仿真研究

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2025-12-19 DOI: 10.1016/j.pnucene.2025.106188

Yanke Wen , Hongyang Chen , Chuangming Ning , Yuhan Zheng , Zhengyang Li , Zhenbing Cai

The high-speed flow of coolant in a nuclear reactor flushes the fuel rods, causing them to undergo flow-induced vibration (FIV). Additionally, tiny metal impurity particles are captured by the grid as the coolant flows, resulting in fretting wear. These foreign objects are primarily swarf generated during the machining of components in the nuclear reactor. Prolonged fretting can lead to localized wear of the fuel rod cladding, potentially resulting in fuel breakage and the release of fission products, thereby compromising the safety of the nuclear power plant. Foreign object wear is currently an important cause of fuel rod failure, yet little research has been done in this area. The small size of foreign particles renders the experimental studies of foreign object wear challenging. The finite element method (FEM) was employed to simulate the positions and quantities of variously shaped foreign object particles captured in the high-temperature pressurized water environment. In this paper, a new experimental method for foreign object wear was proposed, and the behavior of fretting wear between different foreign objects and cladding tubes is simulated by combined experimental and simulation studies. The study focused on analyzing the effects of particle shape and normal force on abrasion performance. The simulation results indicate that some foreign objects become lodged between the rigid support protrusion and the fuel rod cladding tube, remaining in place despite coolant flushing. Experimental data closely match simulation results at low loads.

在核反应堆中，冷却剂的高速流动会冲刷燃料棒，导致燃料棒发生流致振动（FIV）。此外，当冷却剂流动时，微小的金属杂质颗粒被网格捕获，导致微动磨损。这些异物主要是核反应堆部件加工过程中产生的碎屑。长时间的微动会导致燃料棒包壳的局部磨损，可能导致燃料断裂和裂变产物的释放，从而危及核电站的安全。异物磨损是目前燃料棒失效的重要原因，但这方面的研究很少。外来颗粒的小尺寸使得外来物磨损的实验研究具有挑战性。采用有限元方法模拟了高温加压水环境中捕获的各种形状的异物颗粒的位置和数量。本文提出了一种新的异物磨损实验方法，采用实验与仿真相结合的方法模拟了不同异物与包层管之间的微动磨损行为。研究重点分析了颗粒形状和法向力对磨损性能的影响。仿真结果表明，在刚性支撑突出部分和燃料棒包壳管之间存在异物，尽管有冷却液冲洗，但异物仍然存在。实验数据与低负荷下的仿真结果吻合较好。

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引用次数: 0

Research on the influence of gas-ingested flow on internal flow characteristics in reactor coolant pumps 注气流动对反应堆冷却剂泵内部流动特性影响的研究

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2025-12-29 DOI: 10.1016/j.pnucene.2025.106212

Dan Ni, Jinyu Yang, Xiang Yu, Hongchi Zhong, Sheng Lu, Yuquan Zhang, Bo Gao, Zhong Li, Ning Zhang

The reactor coolant pump is essential to nuclear power plant safety. During accidents like LOCA, gas entrainment leads to complex two-phase flow, degrading performance and challenging operational integrity. Current understanding of the associated flow mechanisms remains limited. To analyze the gas-liquid interaction mechanisms in an axial-flow reactor coolant pump under gas-ingested conditions, this study conducted theoretical investigations on the gas-liquid two-phase flow in nuclear reactor coolant pump. Flow characteristics under various air volume fractions (AVF), bubble diameters, and flow conditions were obtained using large-eddy simulation (LES) method with the Mixture multiphase model. Through experimental verification, it was found that the calculation method has an error within 2 % compared to actual results, confirming the accuracy of the model and numerical methods. The research systematically summarizes the performance variation patterns of the reactor coolant pump under two-phase flow conditions. The gas–liquid interaction mechanisms were elucidated through Q-criterion vortex identification, while the modified behavior of the spherical casing vortex under two-phase operation was analyzed. These findings contribute to nuclear engineering advancements by providing fundamental data for preventing severe two-phase flow conditions in reactor coolant pumps.

反应堆冷却剂泵对核电厂的安全至关重要。在LOCA等事故中，气体夹带会导致复杂的两相流动，降低性能并挑战作业完整性。目前对相关流动机制的理解仍然有限。为了分析轴流式反应堆冷却剂泵内气液相互作用机理，本研究对核反应堆冷却剂泵内气液两相流动进行了理论研究。采用混合多相模型，采用大涡模拟（LES）方法获得了不同空气体积分数（AVF）、不同气泡直径和不同流动条件下的流动特性。通过实验验证，发现计算方法与实际结果的误差在2%以内，证实了模型和数值方法的准确性。系统总结了两相流条件下反应堆冷却剂泵的性能变化规律。通过q准则涡识别阐明了气液相互作用机理，并分析了两相工况下球面机匣涡的修正行为。这些发现通过为防止反应堆冷却剂泵中严重的两相流动状况提供基础数据，有助于核工程的进步。

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引用次数: 0

foamForNuclear: A unified OpenFOAM multi-physics platform for nuclear applications foamForNuclear：一个统一的OpenFOAM多物理场核应用平台

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

Progress in Nuclear Energy

Pub Date : 2026-03-01 Epub Date: 2026-01-13 DOI: 10.1016/j.pnucene.2026.106250

Giovanni Nervi , Thomas Guilbaud , Carlo Fiorina , Mathieu Hursin , Alessandro Scolaro

OpenFOAM has been extensively used to develop dedicated physics solvers for nuclear applications, ranging from thermal-hydraulics to structural mechanics and neutronics. However, existing frameworks still lack a general-purpose structure for coupling arbitrary physics across multiple regions and meshes. This work presents foamForNuclear, a new OpenFOAM-based open-source multiphysics architecture that addresses this limitation. Building on the established GeN-Foam and OFFBEAT codes, the platform introduces a flexible, modular structure that enables users to configure simulations with different combinations of physics, solvers, and coupling strategies without source code modifications. The platform architecture, physics libraries, and verification and validation processes are detailed. Its capabilities and applicability to both fission and fusion systems are demonstrated through selected test cases, including the simulation of the ORNL 19-pin sodium-cooled fuel assembly, the sloshing of liquid metals in pool-type reactors under seismic loading, and the thermal performance of a dual-coolant lithium-lead fusion blanket.

OpenFOAM已广泛用于开发核应用的专用物理求解器，范围从热工水力学到结构力学和中子学。然而，现有的框架仍然缺乏一种通用的结构来跨多个区域和网格耦合任意物理。这项工作提出了foamForNuclear，一个新的基于openfoam的开源多物理场体系结构，解决了这个限制。基于已建立的GeN-Foam和OFFBEAT代码，该平台引入了灵活的模块化结构，使用户能够在不修改源代码的情况下配置具有不同物理，求解器和耦合策略组合的模拟。详细介绍了平台架构、物理库以及验证和验证过程。通过选定的测试案例，包括ORNL 19针钠冷却燃料组件的模拟，地震载荷下池式反应堆中液态金属的晃动，以及双冷却剂锂铅聚变包层的热性能，证明了其在裂变和聚变系统中的能力和适用性。

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引用次数: 0