Nuclear Engineering and Design最新文献_第7页

Reactive distillation of rare earth elements via solid–solid reaction for treatment of spent nuclear fuel 通过固-固反应处理乏核燃料的稀土元素反应蒸馏

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-02 DOI: 10.1016/j.nucengdes.2025.114717

Eunsoo Lee, Sang Woon Kwon, Chang Hwa Lee

The sustainable management of spent nuclear fuel (SNF) poses significant challenges, particularly in reducing high-level radioactive waste. To address these issues, high-radiation-generating rare-earth elements (RE), must be converted into stable forms for safe long-term storage in a deep geological repository. This study explores the conversion of RECl₃ (RE = Y, La, Ce, Pr, Nd, and Sm) to their corresponding oxides through reactive distillation via a solid–solid reaction with K₂CO₃. It is crucial for reducing the volume and increasing the safety of geological disposal of nuclear waste. Thermodynamic calculations indicate that the reaction between RECl₃ and K₂CO₃ proceeds favorably without a molten salt, as evidenced by low Gibbs free energy values. Experimentally, the reaction was conducted by mixing RECl₃ with K₂CO₃ in a 1: 2.55 M ratio, followed by heating at 550 °C under 0.9 bar and then at 850 °C under vacuum. X-ray diffraction and scanning electron microscopy analyses results confirm the effective conversion of RECl₃ to high-purity RE oxides. Additionally, experiments using a simulated mixture of RECl₃, reflecting actual SNF composition, yielded the same results, demonstrating that RE oxides can be produced even in mixtures. It further emphasizes the process's applicability to real-world SNF management. The proposed approach can enhance the process efficiency as this method allows the oxidation of RECl₃ and the subsequent separation of byproduct (KCl and CO₂) to be performed within one reactor.

乏核燃料的可持续管理构成了重大挑战，特别是在减少高放射性废物方面。为了解决这些问题，必须将产生高辐射的稀土元素（RE）转化为稳定的形式，以便在深地质储存库中长期安全储存。本研究通过与K2CO3的固-固反应，探讨了rec3 （RE = Y, La, Ce, Pr, Nd, Sm）通过反应精馏转化为相应的氧化物。这对于减少核废料的体积和提高地质处置的安全性至关重要。热力学计算表明，在没有熔盐的情况下，RECl3和K2CO3之间的反应进行得很顺利，这证明了较低的吉布斯自由能值。实验中，以1:2 .55 M的比例将RECl3与K2CO3混合，然后在0.9 bar下550℃加热，然后在850℃真空加热。x射线衍射和扫描电镜分析结果证实了RECl3有效转化为高纯稀土氧化物。此外，使用模拟的RECl3混合物进行的实验，反映了实际的SNF组成，得出了相同的结果，表明即使在混合物中也可以产生稀土氧化物。它进一步强调了该过程对实际SNF管理的适用性。所提出的方法可以提高工艺效率，因为该方法允许在一个反应器内进行RECl3的氧化和随后的副产物（KCl和CO2）的分离。

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引用次数: 0

Random vibration analysis of nuclear power plant structures 核电厂结构随机振动分析

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-31 DOI: 10.1016/j.nucengdes.2025.114698

Gizem Çağlar Dal , Kurtuluş Soyluk

In this study, random vibration analysis of a nuclear power plant building under earthquake loading is performed based on a large-magnitude earthquake of Kobe 1995. A typical nuclear power plant structure widely used in China is selected as a numerical model and modeled as a 3D system. Within the scope of the study, random vibration and deterministic analyses were performed on firm, medium, and soft soils to determine the effects of earthquake motions on nuclear power plant systems. In the study, the theory of random vibration analysis based on the filtered white noise (FWN) ground motion model was utilized and it was intended to determine to what extent the FWN model reflects the real earthquake motion. In addition to soil type, the considered power plant system is analyzed for the ground motions showing near-fault and far-fault characteristics. As a result of the study, it is concluded that the FWN ground motion model used to model earthquake ground motion can be used to consider the effect of real earthquakes. It is also underlined that differences in soil type, fault type and analysis methods affect the results for the considered nuclear power plant structure.

本文以1995年神户大地震为例，对某核电站建筑在地震荷载作用下的随机振动进行了分析。选取国内广泛使用的典型核电站结构作为数值模型，建立三维系统模型。在研究范围内，随机振动和确定性分析进行了坚实，中等和软土，以确定地震运动对核电站系统的影响。本研究利用了基于滤波白噪声（filter white noise， FWN）地震动模型的随机振动分析理论，旨在确定FWN模型在多大程度上反映了真实的地震运动。除土壤类型外，还分析了所考虑的电厂系统的近断层和远断层特征的地震动。研究结果表明，用于模拟地震地震动的FWN地震动模型可以考虑实际地震的影响。本文还强调了土壤类型、断层类型和分析方法的差异会影响所考虑的核电站结构的结果。

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引用次数: 0

Carbon dots-mesoporous silica composites for efficient U(Ⅵ) adsorption and in situ fluorescence monitoring 高效吸附U（Ⅵ）和原位荧光监测的碳点-介孔二氧化硅复合材料

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-31 DOI: 10.1016/j.nucengdes.2025.114740

Yanan Ren , Menghan Tang , Ke Li , Yaorui Li

The development of functional materials capable of simultaneous U(Ⅵ) detection and adsorption remains a critical challenge in nuclear waste management. Herein, we present carbon dots (CDs)-functionalized MCM-41 composites (CA@MCM-41 and AA@MCM-41) that synergistically integrate fluorescent sensing and selective adsorption capabilities. Through hydrothermal synthesis of CDs derived from citric acid (CA) and L-aspartic acid (AA), the composites exhibit U(Ⅵ) adsorption selectivity while preserving ordered mesoporosity. The AA@MCM-41 demonstrates higher adsorption capability for U(Ⅵ) under acidic conditions (pH 1–4) leveraging protonated amine-carboxyl synergy. The CA@MCM-41 exhibits better adsorption capacity in pH 5–6. Crucially, the composites exhibit intrinsic fluorescence-U(Ⅵ) adsorption coupling, where uranyl coordination triggers concentration-dependent quenching. The correlation between fluorescence quenching and U(Ⅵ) adsorption is quantitatively analyzed using the Stern-Volmer-type relationship. The excellent linear fits strongly support a static quenching mechanism dominated by the stable complex formation between the anchored CDs and U(Ⅵ). The demonstrated integration of optical feedback positions these composites as transformative solutions for smart nuclear wastewater treatment systems.

开发能够同时检测和吸附铀（Ⅵ）的功能材料仍然是核废料管理中的一个关键挑战。在此，我们提出了碳点（CDs）功能化的MCM-41复合材料（CA@MCM-41和AA@MCM-41），协同集成了荧光传感和选择性吸附能力。通过水热合成柠檬酸（CA）和l -天冬氨酸（AA）衍生的CDs，复合材料在保持有序介孔的同时表现出U（Ⅵ）的吸附选择性。AA@MCM-41在酸性条件下（pH 1-4）利用质子化胺-羧基协同作用对U（Ⅵ）具有较高的吸附能力。CA@MCM-41在pH值5 ~ 6时表现出较好的吸附能力。关键是，复合材料表现出固有的荧光- u（Ⅵ）吸附偶联，其中铀酰配位触发浓度依赖性猝灭。利用stern - volmer关系式定量分析了荧光猝灭与U（Ⅵ）吸附之间的关系。优异的线性配合有力地支持了由锚定cd和U之间稳定络合物形成主导的静态猝灭机制（Ⅵ）。所展示的光学反馈集成使这些复合材料成为智能核废水处理系统的变革性解决方案。

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引用次数: 0

Modeling and simulation of a free-piston Stirling generator with built-in reactor core for space applications 空间应用内置电抗器芯的自由活塞斯特林发电机的建模与仿真

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114729

Guang Teng , Jianying Hu , Qi Lu , Yanlei Sun , Baifeng An , Ercang Luo

In this paper, a novel Stirling generator is proposed, in which a reactor core with flow channels replaces the original high-temperature heat exchanger, consequently, the heat transfer loop between the reactor and the Stirling engine is eliminated, enabling an integrated design of the reactor and thermoelectric converter. In order to ensure reactor criticality and protect the linear motor from radiation, an enlarged reactor core, a neutron reflector and an acoustic transmission tube are innovated into the engine. A numerical model is established and simulations are carried out to study the influence of the newly introduced part on the performance. The results show that the enlarged core, newly introduced reflector and acoustic transmission tube introduce additional exergy loss. Through design optimization, the overall thermoelectric conversion efficiency of the new configuration exceeds 39 %, which is only 4.84 % lower than that of the traditional. This demonstrates that the free-piston Stirling generator with built-in reactor core could be a feasible technology roadmap for new nuclear power.

本文提出了一种新型的斯特林发电机，用带流道的反应堆堆芯代替原有的高温换热器，从而消除了反应堆与斯特林发动机之间的传热回路，实现了反应堆与热电转换器的一体化设计。为了保证反应堆的临界性能和保护直线电机免受辐射的影响，发动机采用了放大堆芯、中子反射器和声传输管。建立了数值模型，并进行了仿真，研究了新引入零件对性能的影响。结果表明，增大的堆芯、新引入的反射器和声透射管会带来额外的火用损失。通过设计优化，新配置的整体热电转换效率超过39%，仅比传统配置低4.84%。这表明内置反应堆堆芯的自由活塞斯特林发电机可能是一种可行的新核电技术路线图。

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引用次数: 0

On modeling the seismic response of spent nuclear fuel assemblies in vertical dry storage casks 立式干贮存桶中乏燃料组件地震响应模拟研究

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114719

Fady A. Elshazly , Elnaz Seylabi , Lee Joungyeul

Dry storage casks (DSCs) have safely contained spent nuclear fuel (SNF) for decades, including at sites in seismically active regions. However, due to the continued absence of a U.S. deep geologic repository for disposal of SNF, they are now expected to remain on-site in dry storage systems for periods of time much longer than originally anticipated. Therefore, it is important to ensure SNF can continue to meet safety regulations while it continues to be stored on-site until it is disposed of. This paper presents a detailed modeling and analysis framework for a mock-up DSC system, including a high-fidelity finite element model of a surrogate fuel assembly. A multi-step approach is proposed to reduce the computational burden of full-system simulations, wherein an auxiliary model is proposed to approximate the boundary conditions for high-resolution fuel assembly models. Comparative studies under various seismic excitations indicate that the multi-step approach accurately reproduces the salient dynamic features, including system-level peak accelerations and localized stress peaks. This strategy achieves over 70% reduction in computational cost relative to a fully detailed DSC model without significant loss of accuracy. Further parametric analyses with an ensemble of ground motions — from Western and Central-Eastern United States sites — demonstrate that the considered fuel assembly, in the absence of degradation effects, maintains stresses less than 100 MPa and 5% damped spectral accelerations less than 8 g.

干储存桶（dsc）安全地储存乏核燃料（SNF）已经有几十年了，包括在地震活跃地区的站点。然而，由于美国一直缺乏用于处理SNF的深层地质储存库，它们现在预计将在现场干储存系统中停留的时间比原先预期的要长得多。因此，重要的是要确保SNF能够继续满足安全法规，同时继续储存在现场，直到它被处置。本文提出了DSC模型系统的详细建模和分析框架，包括替代燃料组件的高保真有限元模型。为了减少全系统模拟的计算负担，提出了一种多步骤方法，其中提出了一个辅助模型来近似高分辨率燃料组件模型的边界条件。在各种地震激励下的对比研究表明，多步方法可以准确地再现系统级峰值加速度和局部应力峰值等显著的动态特征。与完全详细的DSC模型相比，该策略实现了超过70%的计算成本降低，同时没有显著的准确性损失。对美国西部和中东部地区地面运动的进一步参数分析表明，在没有退化效应的情况下，所考虑的燃料组件保持应力小于100 MPa， 5%阻尼谱加速度小于8 g。

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引用次数: 0

Multistep forecasting of state variables in nuclear power plants using deep learning 基于深度学习的核电厂状态变量多步预测

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114731

Marcelo C. Santos , Bernardo M. Caixeta , Andressa S. Nicolau , Cláudio M.N.A. Pereira , Roberto Schirru

In Nuclear Power Plants (NPPs), most monitoring and diagnostic systems operate based on the principle of Detection and Response (D&R), in which operator actions are triggered only after an anomaly is detected. While effective for real-time monitoring, this approach lacks predictive capability, which is critical for anticipating the evolution of accidents and enhancing operational safety. To address this limitation, this study investigates the use of Deep Learning models for multi-horizon forecasting the temporal behavior of key state variables during normal operation and postulated accident scenarios in nuclear reactors. Two datasets were employed: the LABIHS dataset, composed of simulated time series from a Pressurized Water Reactor (PWR) under a Loss-of-Coolant Accident (LOCA), and the SICA dataset, which contains real operational data from the Angra 1 nuclear power plant. The methodology included data preprocessing and data augmentation using instrumentation noise. Four deep learning architectures were evaluated: Long Short-Term Memory (LSTM), Temporal Convolutional Networks (TCN), Time-series Dense Encoder (TiDE), and Neural Hierarchical Interpolation for Time Series (N-HiTS). These models were trained using a sliding window approach and evaluated across multiple forecasting horizons. Comparative results showed that TCN outperformed LSTM among the classical models, while TiDE and N-HiTS achieved the best overall accuracy and stability across all forecasting horizons. With average MAE values of 1.01 ± 2.39 (LABIHS) and 1.45 ± 1.33 (SICA), these findings confirm the effectiveness of modern Deep Learning architectures for predictive monitoring in nuclear power plant operations.

在核电站（NPPs）中，大多数监测和诊断系统都是基于探测和响应（D&；R）原则运行的，即只有在检测到异常后才触发操作员的操作。虽然这种方法对实时监测是有效的，但它缺乏预测能力，而预测能力对于预测事故的演变和提高运行安全性至关重要。为了解决这一限制，本研究探讨了在核反应堆正常运行和假设事故情景下，使用深度学习模型对关键状态变量的时间行为进行多水平预测。使用了两个数据集：LABIHS数据集，由冷却剂丢失事故（LOCA）下压水堆（PWR）的模拟时间序列组成；SICA数据集，包含来自安格拉1号核电站的真实运行数据。该方法包括使用仪器噪声进行数据预处理和数据增强。评估了四种深度学习架构：长短期记忆（LSTM）、时间卷积网络（TCN）、时间序列密集编码器（TiDE）和时间序列神经分层插值（N-HiTS）。这些模型使用滑动窗口方法进行训练，并在多个预测范围内进行评估。对比结果表明，TCN在经典模型中优于LSTM，而TiDE和N-HiTS在所有预测范围内的总体精度和稳定性最好。平均MAE值为1.01±2.39 （LABIHS）和1.45±1.33 (SICA)，这些发现证实了现代深度学习架构在核电厂运行预测监测中的有效性。

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引用次数: 0

Porous Flow Modeling of Axial Gas Redistribution in Fragmented LWR Fuel Rods using MOOSE 基于MOOSE的碎片化轻水堆燃料棒轴向气体再分布多孔流动模拟

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114677

Chiara Genoni , Kyle A. Gamble , Davide Pizzocri , Fabiola Cappia , Tommaso Bergomi , Chase Christen , Seongtae Kwon

Understanding how gas axially redistributes within fragmented fuel pellets is crucial for predicting the behavior of Light Water Reactor (LWR) fuel rods during accidental scenarios. Specifically, the time scale of this phenomenon plays a fundamental role in determining the progression and hazard of a Loss Of Coolant Accident (LOCA), especially when high burn-up fuel in a severe state of fragmentation is involved. This study presents a Computational Fluid Dynamics (CFD) model developed within the Multiphysics Object-Oriented Simulation Environment (MOOSE) to predict the time-scale of plenum depressurization in fragmented Light-Water Reactor (LWR) fuel rods. The model examines the effects of incorporating non-linearities in the friction term by comparing the results with experimental data. These data were collected from an experiment that employed surrogate fuel rods containing pellets subjected to mechanical and/or thermal loadings. The objective of the experiement was to reproduce various severity of fuel cracking and to investigate the influence of fuel fragmentation on the dynamics of axial gas redistribution. The results of this study indicate that under certain flow regime conditions – determined by the value of an equivalent Reynolds number – accounting for the non-linear friction term in Navier–Stokes equations guarantees better predictions for the time-scale of plenum depressurization. Also, the model enabled the simulation of the plenum pressure decay by assigning distinct permeability values to each pellet instead of a single uniform value. Multiple simulations were run across all possible combinations of pellets’ positions, having each pellet assigned with values of permeability extracted from the experimental data. This allowed to quantify the impact of the considering various non-uniform distributions of permeability on the dynamics of axial gas redistribution. The present work findings enhance the understanding of axial gas transport, and provide valuable insights for the integration of a model for predicting the axial gas redistribution during a LOCA scenario into the BISON fuel performance code.

了解气体在破碎燃料球团内如何轴向再分布，对于预测轻水反应堆（LWR）燃料棒在意外情况下的行为至关重要。具体来说，这种现象的时间尺度在确定冷却剂损失事故（LOCA）的进展和危害方面起着至关重要的作用，特别是当涉及到处于严重破碎状态的高燃耗燃料时。本研究提出了在多物理场面向对象仿真环境（MOOSE）中开发的计算流体动力学（CFD）模型，用于预测碎片化轻水反应堆（LWR）燃料棒充气降压的时间尺度。该模型通过将结果与实验数据进行比较来检验在摩擦项中加入非线性的影响。这些数据收集自一项实验，该实验使用含有颗粒的替代燃料棒，承受机械和/或热负荷。实验的目的是再现不同程度的燃料裂解，并研究燃料破碎对轴向气体再分布动力学的影响。本研究的结果表明，在一定的流型条件下-由等效雷诺数的值决定-考虑Navier-Stokes方程中的非线性摩擦项可以更好地预测充气降压的时间尺度。此外，该模型通过为每个颗粒分配不同的渗透率值而不是单一的均匀值，从而能够模拟充气压力衰减。在所有可能的颗粒位置组合中进行多次模拟，并为每个颗粒分配从实验数据中提取的渗透率值。这可以量化考虑各种不均匀渗透率分布对轴向气体再分布动力学的影响。目前的研究结果增强了对轴向气体输送的理解，并为将LOCA情景下预测轴向气体再分配的模型集成到BISON燃料性能代码中提供了有价值的见解。

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引用次数: 0

Correlation for CHF at all inclinations in rectangular channels with one side heated 矩形通道中各倾角的CHF的相关关系

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114734

Mirza M. Shah

Calculation of CHF (Critical Heat Flux) in rectangular channels with one side heated is needed in applications including nuclear and fusion reactors and cooling of computer chips by boiling liquids. While a well-verified correlation for CHF during horizontal and vertical upflow in such channels is available, there is no well-verified correlation for CHF in other orientations. The present research was done to develop a correlation applicable to all orientations. A new correlation is presented which is in reasonable agreement with data from seven sources. The data include inclinations to the horizontal from 0–316 degree, four fluids (water, R-113, FC-72, and PF-5060), hydraulic equivalent diameter 3.2–23.0 mm, heated equivalent diameter 7.3–281 mm, length to diameter ratio 0.14–83, reduced pressure 0.0045–0.071, mass flux 33–6676 kg/m²s, and inlet quality −0.39 to −0.03. The 463 data points from seven experimental studies are predicted with a MAD (Mean Absolute Deviation) of 23.1 %. The same data were also compared to fourteen other correlations. MAD of those correlations ranged from 58 % - 6523 %. The results of this research are presented and discussed. Suggestions are made for further research, including at very low flow rates for which data are not available.

在核聚变反应堆和沸水冷却计算机芯片等应用中，需要计算一侧受热矩形通道的临界热流密度。虽然在这些通道的水平和垂直向上流动期间，CHF的相关性得到了很好的验证，但在其他方向上，CHF的相关性没有得到很好的验证。本研究的目的是建立一种适用于所有方向的相关性。提出了一种新的相关性，与七个来源的数据基本一致。数据包括水平倾角0-316度，四种流体（水，R-113， FC-72和PF-5060），水力等效直径3.2-23.0 mm，加热等效直径7.3-281 mm，长径比0.14-83，减压0.0045-0.071，质量通量33-6676 kg/m2s，进口质量- 0.39至- 0.03。来自7项实验研究的463个数据点的预测MAD（平均绝对偏差）为23.1%。同样的数据还与其他14种相关性进行了比较。这些相关性的MAD范围为58% - 6523%。本文对研究结果进行了介绍和讨论。提出了进一步研究的建议，包括在没有数据的非常低的流速下。

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引用次数: 0

Momentum equations for scaling analysis of natural circulation loops: Principles and application 自然循环环标度分析的动量方程：原理与应用

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-29 DOI: 10.1016/j.nucengdes.2025.114712

Montanini Marco , Carnevali Sofia , Bestion Dominique , Cottarel Valentin , Rossi Lionel

Developments are in progress at Commissariat á l’énergie atomique et aux Énergies Alternatives (CEA) to use the CATHARE code for scaling analysis. It has been shown that mature system codes can perform more detailed scaling analyses than of analytical methods alone. When an integral effect test facility (IET) is available to simulate a reactor transient and when the code correctly predicts it, a code assisted a-posteriori scaling analysis is able to analyze the origins of distortions and quantify the impact of distortions between the IET facility and the reactor transient simulation. The analysis is now extended to natural convection (NC). Two equations are analyzed: integrated mixed momentum equation (MME), describing mass flow rate, and integrated crossed momentum equation (CME), controlling slip ratio and void fraction in the circuit. A very simple exercise is performed on a loop at reactor scale. The same analysis is run for hypothetical reduced-scale IETs designed using power-to-volume scaling with full-height (PTVS-FH) and three-level-scaling with reduced-height (3LS-RH) approaches. The paper introduces the method developed in CATHARE code based on discretized momentum equations. Nodalizations and boundary conditions which simulate the natural circulation (NC) for a decreasing mass inventory in the circuit are described. Dominant terms of mixture and crossed momentum equation are identified. Distortions of the scaled loops with respect to the reactor-type loop are analyzed and preliminary conclusions are presented.

在化学和化学替代材料委员会（CEA），使用CATHARE代码进行比例分析的工作正在取得进展。成熟的系统代码比单独的分析方法可以进行更详细的尺度分析。当一个积分效应测试设施（IET）可用来模拟反应堆瞬态，并且当代码正确预测它时，代码辅助的后验标度分析能够分析扭曲的来源，并量化IET设施和反应堆瞬态模拟之间的扭曲影响。该分析现已扩展到自然对流（NC）。分析了描述质量流量的积分混合动量方程（MME）和控制电路滑移率和空隙率的积分交叉动量方程（CME）。一个非常简单的练习是在反应堆规模的环路上进行的。同样的分析也适用于采用全高度功率体积缩放（PTVS-FH）和低高度三级缩放（3LS-RH）方法设计的小型IETs。本文介绍了在CATHARE代码中基于离散动量方程开发的方法。描述了电路中模拟自然循环（NC）的节点化和边界条件。确定了混合动量方程和交叉动量方程的主导项。分析了标度环相对于电抗器型环的畸变，并给出了初步结论。

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引用次数: 0

Probabilistic response analysis of skewed distribution of nuclear power plant structural parameters 核电厂结构参数偏态分布的概率响应分析

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-29 DOI: 10.1016/j.nucengdes.2025.114697

Wenfu He , Yuxiang Zhou , Ziduan Shang , Tao Wang , Jiawei Ji

Isolation is widely recognized as one of the most effective technologies for protecting nuclear power structure (NPS) from seismic events. As with NPS, the seismic performance of nuclear plant isolated structure (NPIS) is affected not only by the inherent variability of seismic ground motions but also by significant uncertainties in structural characteristics. The addition of an isolation layer using isolation bearings introduces further uncertainty and raises concerns about the reliability of isolated structures. A probability density evolution method (PDEM) is proposed to evaluate the reliability of NPIS under various operational conditions. First, the NPIS is simplified into a nonlinear model representing the superstructure and the isolation layer, and a probability density evolution analysis procedure (PDEAP) is developed. The analysis considers the uncertainty of the parameters of the superstructure and the isolation layer and uses a Weibull distribution to model the skewed uncertainty. A shaking table test of the NPIS is then conducted. Finally, probability density analyses of the displacement are performed for the base earthquake (OBE), the safe shutdown earthquake (SSE) and the over-design base earthquake (2SSE and 3SSE). The shaking table test results indicate that, compared with the NPS, the NPIS reduces peak displacement by 38.71 %, 50.12 %, and 53.21 % under ground motion amplitudes of 0.3 g, 0.6 g, and 0.9 g, respectively. The probability density evolution analysis further reveals that as the peak ground acceleration (PGA) increases from the OBE to 3SSE, the displacement probability density function (PDF) transitions from a narrow, high peak to a broader, flatter distribution, indicating a significant increase in the variability of structural response.

隔震是公认的保护核电结构免受地震影响的最有效技术之一。与核电厂隔离结构一样，核电厂隔离结构的抗震性能不仅受到地震地震动的固有变异性的影响，而且受到结构特征的显著不确定性的影响。使用隔震轴承增加的隔震层引入了进一步的不确定性，并引起了对隔震结构可靠性的担忧。提出了一种概率密度演化法（PDEM）来评估NPIS在各种运行条件下的可靠性。首先，将NPIS简化为表示上部结构和隔震层的非线性模型，并开发了概率密度演化分析程序（PDEAP）。该分析考虑了上部结构和隔振层参数的不确定性，并采用威布尔分布来模拟偏不确定性。然后对NPIS进行了振动台试验。最后，对基础地震（OBE）、安全停堆地震（SSE）和超设计基础地震（2SSE和3SSE）进行了位移概率密度分析。振动台试验结果表明，与NPS相比，NPIS在0.3 g、0.6 g和0.9 g地震动幅值下，峰值位移分别减少了38.71%、50.12%和53.21%。概率密度演化分析进一步表明，随着峰值地加速度（PGA）从OBE到3SSE的增加，位移概率密度函数（PDF）从窄、高的峰值转变为更宽、更平坦的分布，表明结构响应的变异性显著增加。

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