Nuclear Engineering and Design最新文献_第4页

Validation of NekRS pebble bed modeling for low flow scenarios 低流量情景下NekRS卵石床模型的验证

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.2025.114693

Michael Seneca , Haomin Yuan , Tri Nguyen , David Reger , Victor Coppo Leite , Luiz Aldeia , Joseph Seo , Seth Macias , Dezhi Dai , Elia Merzari , Yassin Hassan

This study validates the high-fidelity NekRS Large Eddy Simulation (LES) framework for modeling conjugate heat transfer in low-flow pebble bed reactor scenarios, using the SANA experiments as a benchmark. Verification began by comparing the MOOSE Net Radiation Method to a 1D analytical model for a tall-cavity case, yielding Nusselt numbers within 1% relative difference. Next, the NekRS P1 approximation was compared to MOOSE in a single-pebble case, where temperature and velocity profiles agreed within 5% relative difference. Validation proceeded with the 283-pebble bed experiments at Texas A&M, in which NekRS velocity profiles fairly matched experimental measurements despite uncertainties in pebble placements. Finally, full conjugate heat transfer simulations of the SANA pebble beds (5 kW helium–graphite and helium–alumina cases) were performed, with the optimized configuration producing a mean temperature error of 1.56%. The effective chamfer thermal conductivity (representing conductive contact between pebbles) introduced significant uncertainty in modeling the SANA experiments; further work will optimize this parameter based on a refined contact resistance model. Simulations will also be extended to higher-power conditions to continue validating conjugate heat transfer in NekRS for pebble-bed reactor designs.

本研究以SANA实验为基准，验证了高保真NekRS大涡模拟（LES）框架在低流量球床反应器场景下的共轭传热建模。通过将MOOSE净辐射方法与高腔情况下的一维分析模型进行比较，验证开始，得到的努塞尔数相对差异在1%以内。接下来，将NekRS P1近似与MOOSE在单卵石情况下进行比较，温度和速度曲线的相对差异在5%以内。在德克萨斯a&m进行的283个卵石床实验中，NekRS的速度剖面与实验测量结果相当吻合，尽管卵石的位置存在不确定性。最后，对SANA球床（5 kW氦-石墨和氦-氧化铝）进行了全共轭传热模拟，优化后的球床平均温度误差为1.56%。有效倒角导热系数（表示鹅卵石之间的导电接触）在SANA实验建模中引入了显著的不确定性；进一步的工作将基于改进的接触电阻模型对该参数进行优化。模拟还将扩展到更高功率的条件下，以继续验证NekRS中用于球床反应器设计的共轭传热。

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引用次数: 0

Experimental investigations of Pool hydrodynamics and aerosol removal under low momentum injection 低动量注入池流体力学与气溶胶去除实验研究

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.2025.114735

Nabil Ghendour , Detlef Suckow , Abdelouahab Dehbi , Michael Klauck

A detailed experimental database was generated from hydrodynamic characterization a Wire Mesh Sensor (WMS) and a High-Speed Camera (HSC) in the TRISTAN facility and aerosol retention measurements in the SAAB facility, performed under identical injection conditions in the low-momentum globule regime. Nitrogen was injected through 5 mm and 10 mm nozzles, spanning a flow range of 1–10 l_n/min, with a water submergence of 300 mm. This corresponds to a Weber number up to 7.3. HSC images showed aperiodic globule formation, revealing intensified bubble coalescence and break-up with increasing flow rates. The image processing demonstrated a high Gas Void Fraction (GVF) in the injection region, which increases with flow rate. The WMS data were collected at heights of 40, 100, and 200 mm above the nozzle tip. The data analysis mirrored GVF trends from the HSC images. Furthermore, the velocity profile of the gas phase was analyzed. Notably, the impact of the nozzle diameter is prominent in the injection region near the nozzle. An advanced algorithm was developed to track and extract globule formation characteristics from HSC images. A new scaling concept to describe globule characteristics as function of the Weber number was introduced and validated using additional experimental data. Based on this, a new correlation for the globule diameter is proposed for Weber number up to 70. Corresponding aerosol pool scrubbing tests were conducted and showed that particle retention is roughly insensitive to the gas flow rate within the experimental range, but is enhanced as particle inertia increases. In addition, use of the smaller nozzle under similar flow rates results mostly in a slight improvement of aerosol removal. These high-fidelity data can serve to develop and validate CFD models for hydrodynamics and aerosol pool scrubbing.

在相同的注入条件下，通过TRISTAN设施的金属丝网传感器（WMS）和高速相机（HSC）和SAAB设施的气溶胶保留测量，生成了详细的实验数据库。氮气通过5毫米和10毫米的喷嘴注入，流量范围为1-10 ln/min，水深为300 mm。这相当于韦伯数达到7.3。HSC图像显示非周期性的球状形成，随着流量的增加，气泡的合并和破裂加剧。图像处理表明，注入区气含率（GVF）较高，且随流量增大而增大。WMS数据分别在喷嘴尖端上方40、100和200 mm处采集。数据分析反映了HSC图像的GVF趋势。进一步分析了气相的速度分布。值得注意的是，喷嘴直径的影响在喷嘴附近的注射区域是突出的。开发了一种先进的算法来跟踪和提取HSC图像中的球形成特征。引入了一个新的尺度概念，将球的特征描述为韦伯数的函数，并使用额外的实验数据进行了验证。在此基础上，提出了韦伯数≤70时球直径的新相关性。进行了相应的气溶胶池洗涤试验，结果表明，在实验范围内，颗粒滞留对气体流速基本不敏感，但随着颗粒惯性的增加而增强。此外，在相同的流量下，使用较小的喷嘴，大多数情况下气溶胶去除效果略有改善。这些高保真数据可以用于开发和验证流体动力学和气溶胶池洗涤的CFD模型。

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引用次数: 0

High-fidelity neutron transport simulation for pebble-bed HTR in HNET HNET中球床高温堆高保真中子输运模拟

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.2025.114728

Chen Hao , Yuchen Wen , Yizhen Wang

HNET is a high-fidelity neutron transport program developed for 3D reactor core simulation. To enhance its high-fidelity neutron transport simulation capability for pebble-bed HTR, a three-dimensional Method of Characteristics (3D-MOC) solver with Linear Source Approximation (LSA) is developed in HNET recently. The present work gives a comprehensive description on this newly developed solver with emphasizes on its calculation scheme, geometric modelling capability, parallel strategy and acceleration techniques. Computational efficiency of this 3D-MOC solver is enhanced by using virtual mesh CMFD method (vcCMFD) and modular track arrange strategy. This pebble-bed HTR 3D-MOC solver in HNET is verified with a pebble-bed HTR whole core model developed on HTR-10 benchmark. Results show that it takes about 189 core-hours to perform whole-core criticality simulation, which demonstrates its feasibility and efficiency for pebble-bed HTR high-fidelity neutron transport simulation.

HNET是为三维反应堆堆芯模拟而开发的高保真中子输运程序。为了提高球床高温堆的高保真中子输运模拟能力，最近在HNET中开发了一种基于线性源近似的三维特征法（3D-MOC）求解器。本文对这种新开发的求解器进行了全面的描述，重点介绍了它的计算方案、几何建模能力、并行策略和加速技术。采用虚拟网格CMFD方法（vcCMFD）和模块化轨迹排列策略，提高了三维moc求解器的计算效率。用基于HTR-10基准开发的球床HTR全岩心模型对HNET中的球床HTR 3D-MOC求解器进行了验证。结果表明，全堆临界模拟所需时间约为189芯小时，证明了球床高温堆高保真中子输运模拟的可行性和有效性。

引用次数: 0

Thermal evolution in molten salt conditioned waste drums 熔盐条件下废桶的热演化

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.2025.114742

Suresh Seetharam , Quoc Tri Phung , Vojtech Galek , Anna Sears , Lander Frederickx , Eduardo Ferreira

This paper presents an experimental and numerical investigation of the thermal evolution of a conditioned molten salt oxidation (MSO) residue contained within 100 L drums. Three types of binders were employed: (i) alkali-activated material (AAM) with a metakaolin precursor (AAM_MK), (ii) AAM with a blast furnace slag (BFS) precursor, and (iii) a blended cement mix, each with varying waste loadings. The study primarily involved isothermal and semi-adiabatic calorimetry experiments to develop a comprehensive dataset of hydration curves, which serve as direct inputs for a heat transfer model. Drum-scale experiments on the reconditioned MSO residue waste with metakaolin precursor were successfully designed and executed. Thermal evolution within the drum was monitored using thermal sensors strategically placed at various locations. A standard heat transfer model was employed for blind predictions of thermal evolution within the drum. Calorimetric measurements of the different waste forms indicated that the addition of MSO residue delayed hydration and geopolymerization in both the cementitious and alkali-activated matrices. The numerical model reasonably captured the primary features of thermal evolution, particularly the peak measured temperature data (> 80 °C) at the core of the drum conditioned with the AAM_MK binder, while also highlighting the uncertainty in the sensitive model parameters. It is anticipated that for typical drum sizes exceeding 200 L used in pre-disposal storage, peak temperatures could surpass 100 °C. Consequently, further studies on the long-term stability of reconditioned waste forms exposed to high early age temperatures are warranted.

本文对100 L圆筒内条件熔盐氧化（MSO）残渣的热演化进行了实验和数值研究。使用了三种类型的粘合剂：(i)碱活化材料（AAM）与偏高岭土前驱体（AAM_MK），（ii） AAM与高炉炉渣（BFS）前驱体，以及（iii）混合水泥混合物，每种都具有不同的废物负荷。该研究主要涉及等温和半绝热量热实验，以建立一个综合的水化曲线数据集，作为传热模型的直接输入。设计并成功地进行了以偏高岭土为前驱体的MSO废渣改造后的转鼓试验。使用放置在不同位置的热传感器监测鼓内的热演变。采用标准传热模型对鼓内热演化进行了盲预测。不同废物形式的量热测量表明，MSO残渣的加入延迟了胶凝和碱活化基质中的水化和地聚合。该数值模型合理地捕捉了热演化的主要特征，特别是在AAM_MK粘结剂调节下的鼓芯处的峰值测量温度数据（> 80°C），同时也突出了敏感模型参数的不确定性。预计在预处理存储中使用的超过200升的典型桶尺寸，峰值温度可能超过100°C。因此，有必要进一步研究暴露于早期高温下的修复废物形式的长期稳定性。

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引用次数: 0

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