Progress in Nuclear Energy最新文献_第3页

Advancements to generalized equivalence theory for preserving cross-sections of whole core simulations 用于保存整个岩心模拟截面的广义等价理论的进展

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-17 DOI: 10.1016/j.pnucene.2024.105504

The conventional two-step approach, without the use of an equivalence method, can introduce significant error for the simulation of non-LWRs (Light Water Reactors), especially fast reactors, due to the large leakage and high anisotropic neutron distribution. To improve upon the accuracy of the two-step approach, a 3D whole core model is simulated with a transport method to generate region-wise homogenized cross-sections (XS). These XS can then be used in a diffusion whole core solver during the second step to extend the application to cycle and transient analysis. Discontinuity factors (DFs) are then introduced to improve the accuracy during the simulation of the second step. With properly generated DFs from Generalized Equivalence Theory (GET), the region-averaged solutions from the first transport step can then be reproduced by the second step diffusion solver. The Monte Carlo method was selected to perform the whole core transport simulation to generate region-wise XS. However, simulating whole core problems with Monte Carlo may result in poor statistics near the peripheral region especially for partial current tallies. This paper introduces an advancement to GET to reproduce region-wise solutions for select regions when the reaction rates and surface currents have good statistics in these regions but poor statistics in other regions. A reference high-fidelity model was constructed using the Serpent 2 Monte Carlo code based on the EBR-II benchmark evaluation and verification was carried out using the TriPEN-4 method in PARCS. The results show that it is possible to reproduce the exact eigenvalue and power distributions of whole core problems in a feasible manner.

由于存在大量泄漏和各向异性的中子分布，不使用等效方法的传统两步法会给非轻水反应堆（尤其是快堆）的模拟带来很大误差。为了提高两步法的精确度，采用传输方法对三维全堆芯模型进行模拟，以生成区域均匀化截面（XS）。在第二步中，这些 XS 可用于扩散整个岩心求解器，从而将应用扩展到循环和瞬态分析。然后引入不连续因子 (DF)，以提高第二步模拟的精度。有了根据广义等效理论（GET）正确生成的 DF，第二步扩散求解器就可以再现第一步传输过程中的区域平均解。我们选择蒙特卡洛方法来进行整个岩心的传输模拟，以生成区域 XS。然而，用蒙特卡罗法模拟整个岩心问题可能会导致外围区域附近的统计数据较差，尤其是部分电流统计。本文介绍了 GET 的一种改进方法，即当反应速率和表面电流在选定区域的统计量较好，而在其他区域的统计量较差时，在这些区域重现区域解。在 EBR-II 基准评估的基础上，使用 Serpent 2 蒙地卡罗代码构建了一个参考高保真模型，并使用 PARCS 中的 TriPEN-4 方法进行了验证。结果表明，有可能以可行的方式再现整个核心问题的精确特征值和功率分布。

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引用次数: 0

Rotor fault diagnosis of centrifugal pumps in nuclear power plants based on CWGAN-GP-CNN for imbalanced dataset 基于不平衡数据集的 CWGAN-GP-CNN 核电站离心泵转子故障诊断

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-16 DOI: 10.1016/j.pnucene.2024.105500

As a crucial device in nuclear power plants, centrifugal pumps undertake the critical role of cooling water circulation. Centrifugal pump rotor misalignment and unbalanced faults cause pump performance degradation, vibration increase, and equipment damage, thus seriously affecting the safety and reliability of nuclear power plants. In the process of centrifugal pump rotor fault, the difficulty in obtaining data samples and the limited amount of data can lead to an imbalance problem between the quantity of normal state and fault state samples in the dataset. In order to solve the problem, this paper proposed a CWGAN-GP model for generating rotor fault data based on CGAN and WGAN-GP models, and combined it with a two-stream CNN model to realize the rotor fault diagnosis with an imbalanced dataset. The quality and performance of the data generated by the proposed method were evaluated and validated in terms of visualization analysis, statistical indicators, and comparison with different data generation models. The results show that the CWGAN-GP model can generate high-quality data. Meanwhile, compared with other models on datasets with different degrees of imbalance, the two-stream CNN model is more effective in fault diagnosis on the expanded dataset by the CWGAN-GP model, and the improvement of fault diagnosis accuracy ranges from 1.40% to 13.33%.

作为核电站的重要设备，离心泵承担着冷却水循环的关键作用。离心泵转子不对中和不平衡故障会导致泵性能下降、振动加剧和设备损坏，从而严重影响核电站的安全性和可靠性。在离心泵转子故障处理过程中，由于数据样本获取困难，数据量有限，会导致数据集中正常状态样本与故障状态样本数量不平衡的问题。为了解决这一问题，本文在 CGAN 和 WGAN-GP 模型的基础上，提出了生成转子故障数据的 CWGAN-GP 模型，并将其与双流 CNN 模型相结合，实现了不平衡数据集下的转子故障诊断。从可视化分析、统计指标以及与不同数据生成模型的比较等方面对所提出方法生成的数据的质量和性能进行了评估和验证。结果表明，CWGAN-GP 模型可以生成高质量的数据。同时，在不同失衡程度的数据集上，与其他模型相比，CWGAN-GP 模型的双流 CNN 模型在扩展数据集上的故障诊断效果更好，故障诊断准确率提高了 1.40% 至 13.33%。

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引用次数: 0

Design and optimization of the components of a molten salt, thorium-fueled accelerator driven system 熔盐钍燃料加速器驱动系统组件的设计与优化

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-15 DOI: 10.1016/j.pnucene.2024.105486

Molten Salt Accelerator-Driven Systems (MoSTADS) have been attracting a lot of research interest lately due to their unique characteristics and advantages, including reduced radiation damage of the fuel, and stable operation achieved through online fuel feeding process. Simulations of an experimental molten salt test facility being developed in the Thermal Hydraulic Research Laboratory (THRL) at Texas A&M University, were conducted using Monte Carlo radiation transport methods, to design and optimize selected components of the system. The system consists of a proton beam generated by an accelerator, impinging on a target to generate neutrons, which can be used induce fission reactions within a thorium fueled, high-temperature molten salt forced convection test loop. Parametric studies were performed to optimize several key components of the system including target material, proton beam energy, target thickness and location, and reflector thickness. Furthermore, in order to ensure the safe operation of the facility, parametric studies were also performed to identify the composition and thickness of the system shielding that would be needed to satisfy acceptable exposure limits.

熔盐加速器驱动系统（MoSTADS）因其独特的特性和优势，包括减少燃料的辐射损伤以及通过在线燃料供给过程实现稳定运行等，近来吸引了大量研究人员的关注。德克萨斯A&M大学热液压研究实验室（THRL）正在开发一种实验性熔盐试验设备，我们采用蒙特卡洛辐射传输方法对该设备进行了模拟，以设计和优化系统的选定组件。该系统由加速器产生的质子束组成，质子束撞击目标产生中子，中子可用于诱导以钍为燃料的高温熔盐强制对流试验回路中的裂变反应。参数研究旨在优化系统的几个关键组件，包括靶材料、质子束能量、靶厚度和位置以及反射器厚度。此外，为了确保设施的安全运行，还进行了参数研究，以确定满足可接受的暴露限值所需的系统屏蔽成分和厚度。

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引用次数: 0

Modeling and benchmarking XRF analysis using MCNP for applications in accident tolerant fuel and cladding 利用 MCNP 对 XRF 分析进行建模和基准测试，以应用于耐事故燃料和包壳

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-14 DOI: 10.1016/j.pnucene.2024.105487

There is an interest in using nondestructive testing (NDT) methods for the preliminary investigation of accident-tolerant fuel cladding materials, such as chromium (Cr) coated Zircaloy-4 (Zr4). One promising application is X-ray fluorescence (XRF) analysis. Computational methods, such as Monte Carlo N-Particle Transport (MCNP) 6.2, can be used to expand algorithms based on XRF measurements, however, it has been demonstrated that MCNP is more sensitive to modeling imperfections at lower energies (

\leq

80 keV). In this work, several MCNP models were developed to evaluate the XRF measurements given by a Niton XL-5 device to minimize deviations at low energies. The final model was benchmarked to an experimental XRF measurement of Cr-coated Zr4 taken by the XL-5. The percent error in the resulting XRF peak intensities was within

\pm

4.92% for the K_α1 peaks and within

\pm

16.0% for the K_β1. The discrepancies in the magnitude of these errors are largely due to the K_β1 peaks having far fewer counts in the spectra that were compared. Nonetheless, these results demonstrate the potential for MCNP 6.2 to accurately predict low-energy X-ray interactions such as XRF. The deviations observed were similar to those seen in the 0.1–1 MeV range in prior works, despite only being in the

\sim

5–30 keV range themselves.

人们对使用无损检测（NDT）方法来初步调查耐事故燃料包壳材料（如铬（Cr）涂层锆合金-4（Zr4））很感兴趣。X 射线荧光 (XRF) 分析是一项很有前景的应用。Monte Carlo N-Particle Transport (MCNP) 6.2 等计算方法可用于扩展基于 XRF 测量的算法，然而，事实证明 MCNP 在较低能量（≤ 80 keV）时对建模缺陷更为敏感。在这项工作中，开发了多个 MCNP 模型来评估由 Niton XL-5 设备提供的 XRF 测量结果，以尽量减少低能量时的偏差。最终模型以 XL-5 对镀有铬的 Zr4 进行的 XRF 实验测量为基准。得出的 XRF 峰强度百分比误差，Kα1 峰在± 4.92% 以内，Kβ1 峰在± 16.0% 以内。这些误差幅度上的差异主要是由于在比较的光谱中，Kβ1 峰的计数要少得多。尽管如此，这些结果还是证明了 MCNP 6.2 在准确预测低能 X 射线相互作用（如 XRF）方面的潜力。所观察到的偏差与之前研究中 0.1-1 MeV 范围内的偏差相似，尽管其本身仅在∼ 5-30 keV 范围内。

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引用次数: 0

Large-eddy simulation of turbulent flow and heat transfer of helically corrugated tubes in the intermediate heat exchanger of a very-high-temperature gas-cooled reactor 超高温气冷式反应堆中间热交换器中螺旋波纹管的湍流和传热的大涡流模拟

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-14 DOI: 10.1016/j.pnucene.2024.105488

The intermediate heat exchanger (IHX) is a vital component of very-high-temperature gas-cooled reactors (VHTRs) utilized for thermal applications of nuclear energy, specifically for hydrogen production. Enhancing the heat transmission capacity of IHXs is essential to provide sufficient heat for thermal processes. This study uses large eddy simulations to investigate IHX models consisting of both a smooth circular tube and helically corrugated tubes with five different geometric parametrizations. The results show how turbulent flow and heat transfer depend on the geometric parameters. Based on theories such as boundary layer theory, field synergy, and extreme dissipation, the characteristics of boundary layer separation, secondary flow, turbulent transport, field synergy, and dissipation characteristics in helically corrugated tubes are quantitatively analyzed. The study also investigates enhanced heat transfer mechanisms within the helically corrugated tubes, and the results attribute the enhanced heat transfer in helically corrugated tubes to the helical structure, which hinders the development of a fluid boundary layer, strengthens the intensity of secondary flow and turbulent transport, improves the synergy between the fluid velocity field and the temperature gradient, and reduces the thermal potential energy loss during the fluid heat transfer.

中间热交换器（IHX）是用于核能热应用（特别是制氢）的超高温气冷堆（VHTR）的重要组成部分。提高 IHX 的热传输能力对于为热过程提供足够的热量至关重要。本研究利用大涡流模拟研究了由光滑圆管和螺旋波纹管组成的 IHX 模型，其中螺旋波纹管有五种不同的几何参数。结果显示了湍流和传热如何取决于几何参数。基于边界层理论、场协同和极端耗散等理论，定量分析了螺旋波纹管中的边界层分离、二次流、湍流传输、场协同和耗散特性。研究还探讨了螺旋波纹管内的强化传热机理，结果表明螺旋波纹管内的强化传热是由于螺旋结构阻碍了流体边界层的形成，加强了二次流和湍流输运的强度，改善了流体速度场和温度梯度之间的协同作用，减少了流体传热过程中的热势能损失。

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引用次数: 0

Thermal-hydraulic analysis of helical coil once-through steam generators under complex oceanic conditions 复杂海洋条件下螺旋线圈直通式蒸汽发生器的热工水力分析

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105484

The helical coil once-through steam generator (HCOTSG) possesses several advantages, including excellent thermal efficiency, compact size, rapid start-up, and low energy consumption. These qualities make it particularly well-suited for use in ship reactors. This paper introduces a one-dimensional thermal-hydraulic calculation program for helical coil once-through steam generators (HCOTSGs) operating under complex oceanic conditions, utilizing the finite difference method alongside fixed boundary conditions and staggered grid techniques. By incorporating the heat transfer and friction relations to the helical coil, as well as considering the geometric structure of the helical coil when incorporating ocean conditions, the characteristics of the helical coil are reflected in the one-dimensional code. The impact of oceanic conditions is considered by modifying the momentum equation. This paper considers highly complex oceanic conditions that involve a coupling of inclining and oscillation. Furthermore, the accuracy of the developed program is verified by comparing the HCOTSG design parameters of the MRX reactor with the calculated values from internationally recognized prediction programs against the findings presented here.

螺旋盘管直通式蒸汽发生器（HCOTSG）具有多个优点，包括热效率高、体积小、启动快和能耗低。这些优点使其特别适合用于船用反应堆。本文采用有限差分法、固定边界条件和交错网格技术，介绍了在复杂海洋条件下运行的螺旋线圈直通式蒸汽发生器（HCOTSG）的一维热工水力计算程序。通过将传热和摩擦关系纳入螺旋线圈，以及在纳入海洋条件时考虑螺旋线圈的几何结构，螺旋线圈的特性在一维代码中得到了反映。通过修改动量方程来考虑海洋条件的影响。本文考虑了涉及倾斜和振荡耦合的高度复杂的海洋条件。此外，通过将 MRX 反应堆的 HCOTSG 设计参数与国际公认的预测程序的计算值进行比较，验证了所开发程序的准确性。

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引用次数: 0

Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Oxygen embrittlement Zircaloy-4 燃料销在模拟失壳事故条件下失效：氧脆性

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105485

An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates.

为了研究印度压水重水反应堆（PHWR）燃料销在模拟失去冷却剂事故（LOCA）条件下的氧脆性，进行了广泛的实验研究。锆合金燃料包壳在使用和 LOCA 条件下同时经历蠕变和腐蚀。对 Zircaloy-4 燃料栓进行预氧化处理，以达到不同的氧化层厚度，从而实现在役条件。然后在蒸汽环境中对这些预氧化管进行爆裂试验，以模拟 LOCA 情况。本研究旨在通过考虑瞬态加热过程中氧化和变形对 LOCA 域中包层行为的交叉影响，加深对氧化对 LOCA 情景下包层微观结构和燃料引脚机械响应的影响的理解。使用 FESEM、XRD 和拉曼光谱对爆前和爆后样品中的氧化层形态进行了研究。在某些情况下，当爆燃试验期间燃料销在蒸汽中加热时，内部氧化层的生长速度快于外部氧化层。在瞬态加热过程中，氧化层径向和圆周裂纹的增长以及分层的发生都加快了氧气和氢气的吸收。爆破前和爆破后样品的氢吸收与氧吸收有关。在氧化前的样品中，氢浓度随氧浓度的增加而增加。由于形成了保护性氧化层，爆破后的样品中氧气和氢气浓度较小。高温氧化层是在极高的加热速率下形成的。

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引用次数: 0

Derivative dynamic time warping algorithm with introduced correction for varying load fault diagnosis of nuclear power system steam turbine units 用于核电系统蒸汽轮机组不同负荷故障诊断的导入校正的衍生动态时间扭曲算法

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-13 DOI: 10.1016/j.pnucene.2024.105490

The Derivative Dynamic Time Warping (DDTW) algorithm is improved to address the multi-parameters time series classification problem faced by nuclear power system steam turbine units varying load fault diagnosis. Firstly, the entire load changing process is treated as a single sample rather than multiple time-step-samples. This ensures the complete information on the load changing processes, while avoiding interference from normal data fluctuations during faults. Secondly, Time Series Position Coefficient and Time Series Length Coefficient are proposed to correct the DDTW algorithm from two perspectives: the sequences lengths and the data positions in the sequences. This solves the singularities and timeline scaling problems, thereby preventing interference introduced by data sequences' lengths differences and ''similar data appearing at different times'' problem. The nuclear power system steam turbine unit simulation model was built to obtain load changing processes data under normal and faults statuses. In the varying load fault diagnosis test based on these data, the improved DDTW algorithm achieved an accuracy of 1.38%–12.06% higher than other methods, reaching 87.50%. Finally, The Deep Convolutional Generative Adversarial Networks (DCGAN) model was used to generate data to supplement the limited samples of complete load changing processes, and the accuracy of the novel method increased to 95.51% with the increase of data used to support the comparison.

针对核电系统汽轮机组变化负荷故障诊断所面临的多参数时间序列分类问题，对衍生动态时间扭曲（DDTW）算法进行了改进。首先，整个负荷变化过程被视为一个样本，而不是多个时间步长样本。这确保了负荷变化过程的完整信息，同时避免了故障期间正常数据波动的干扰。其次，提出了时间序列位置系数和时间序列长度系数，从序列长度和序列中的数据位置两个角度对 DDTW 算法进行修正。这解决了奇异性和时间轴缩放问题，从而防止了数据序列长度差异带来的干扰和 "相似数据出现在不同时间 "问题。建立了核电系统蒸汽轮机组仿真模型，以获取正常和故障状态下的负荷变化过程数据。在基于这些数据的负荷变化故障诊断测试中，改进后的 DDTW 算法的准确率比其他方法高出 1.38%-12.06%，达到 87.50%。最后，使用深度卷积生成对抗网络（DCGAN）模型生成数据，以补充完整负载变化过程的有限样本，随着用于支持比较的数据的增加，新方法的准确率提高到 95.51%。

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引用次数: 0

Interaction behaviour of alloy 690 upon exposure to P2O5 containing borosilicate glass at simulated vitrification conditions 合金 690 在模拟玻璃化条件下暴露于含 P2O5 的硼硅玻璃时的相互作用行为

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-11 DOI: 10.1016/j.pnucene.2024.105480

This study investigates the interaction between alloy 690 and two types of glasses: pristine borosilicate and P₂O₅-containing borosilicate glass, at typical glass pouring temperatures encountered during the vitrification of high-level radioactive waste in the back-end of Nuclear Fuel Cycle (NFC). Partial crystallization of both glasses was observed at the alloy 690/glass interface, with certain, though not entirely identical, crystalline phases forming at the interface. The density of these crystalline phases is significantly higher than that of the surrounding glass, which raises concerns about these reaction products settling at the bottom of the furnace. Such sedimentation could potentially obstruct the freeze valve, thereby halting the vitrification process. Additionally, intergranular grooves on the alloy surface exposed to P₂O₅-containing borosilicate glass were found to disappear with prolonged exposure. This phenomenon is attributed to the strong corrosive action of the highly basic and oxidizing P₂O₅-bearing glass, leading to the peeling away of the entire exposed surface of alloy 690.

本研究调查了在核燃料循环（NFC）后端高放射性废物玻璃化过程中遇到的典型玻璃浇注温度下，合金 690 与两种玻璃（原始硼硅酸盐玻璃和含 P2O5 的硼硅酸盐玻璃）之间的相互作用。在合金 690/玻璃的界面上观察到了两种玻璃的部分结晶，在界面上形成了某些结晶相，但并不完全相同。这些结晶相的密度明显高于周围玻璃的密度，这让人担心这些反应产物会沉淀在熔炉底部。这种沉淀可能会阻塞冻结阀，从而停止玻璃化过程。此外，在长期暴露于含 P2O5 的硼硅玻璃中的合金表面上，晶间凹槽会消失。这种现象归因于高碱性和氧化性 P2O5 玻璃的强烈腐蚀作用，导致合金 690 的整个暴露表面剥落。

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引用次数: 0

Study of the effect of seawater properties on the performance of molten fuel fragmentation 研究海水特性对熔融燃料破碎性能的影响

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-10 DOI: 10.1016/j.pnucene.2024.105483

In this study, in order to explore the effects of seawater properties on the fragmentation behavior of melt jets during a severe core meltdown accident in the light water reactor, visualized fragmentation experiments are carried out by releasing superheated melt into subcooled water at different coolant salinities, melt temperatures, water temperatures, and melt penetration velocities using the VTMCI (Visualized Thermo-hydraulic characteristics in Melt Coolant Interaction) facility at Sun Yat-Sen University. It is found that under the current experimental conditions, as the coolant salinity increases, the size of the debris decreases, while the variation of debris sphericity and debris bed porosity is insignificant. When the water temperature or melt temperature increases, the sphericity of the debris is higher, and the porosity of the debris bed and the size of the debris decrease. When the penetration rate of the melt is higher, smaller particles can be generated, but it has no significant impact on the debris bed porosity and debris sphericity. The Weber number theory can be used to predict the median diameter of debris, while the Stephan number (St) can be used to predict the trends in debris bed porosity and debris sphericity. This study contributes to a deeper understanding of the actual fragmentation process mechanism of molten materials in severe accidents of light water reactors. The experimental data obtained will also contribute to the development, validation, and improvement of relevant physical models in China's pressurized water reactor severe accident analysis codes.

本研究利用中山大学的 VTMCI（熔体冷却剂相互作用中的可视化热工水力特性）设备，在不同冷却剂盐度、熔体温度、水温和熔体穿透速度下，将过热熔体释放到过冷水中，进行了可视化碎裂实验，以探索海水特性对轻水堆严重堆芯熔毁事故中熔体喷射碎裂行为的影响。研究发现，在目前的实验条件下，随着冷却剂盐度的增加，碎片的尺寸减小，而碎片球度和碎片床孔隙率的变化不大。当水温或熔体温度升高时，碎片的球度增大，碎片床孔隙率和碎片尺寸减小。当熔体的渗透率较高时，会产生较小的颗粒，但对碎屑床孔隙率和碎屑球度的影响不大。韦伯数理论可用于预测碎片的中值直径，而斯蒂芬数（St）可用于预测碎片床孔隙率和碎片球度的变化趋势。这项研究有助于加深对轻水反应堆严重事故中熔融材料实际破碎过程机理的理解。所获得的实验数据也将有助于中国压水堆严重事故分析代码中相关物理模型的开发、验证和改进。

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引用次数: 0