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Machine learning-enhanced dynamic probabilistic safety assessment for station blackout in Bushehr nuclear power plant 基于机器学习的布什尔核电站停电动态概率安全评估
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-12 DOI: 10.1016/j.pnucene.2026.106240
Mehdi Yarizadeh-Bene , Mahdi Zangian , Abdolhamid Minuchehr , Hamed Kargaran
Dynamic Probabilistic Safety Assessment (D-PSA) faces significant computational challenges when simulating complex accident scenarios with high-fidelity thermal-hydraulic codes, necessitating innovative approaches to balance accuracy and efficiency. This study introduces a machine learning-enhanced Dynamic Event Tree (DET) module that improves risk analysis quantification by integrating deep learning and Support Vector Machine (SVM). This module utilizes a FORTRAN-based submodule for automated generation and parallel execution of thermal hydraulic model input files, enabling concurrent simulation of several of dynamic scenarios while a random sampling strategy ensures comprehensive coverage of failure sequences with minimized training data requirements. In this research a Station blackout (SBO), as initiating event at Bushehr nuclear power plant (BNPP) is considered for thermal-hydraulic model benchmark. Consequently, dynamic effect of diesel generators recovery in different time steps are evaluated. As a result, SBO turns into the loss of offsite power (LOOP) accident. The proposed module investigates LOOP accident for comprehensive risk analysis at BNPP. The results show that the machine learning architecture developed achieves good predictive performance, surpassing at least 97 % accuracy in classifying core damage states and reducing scenario evaluation time. High-resolution dynamic modeling combined with computational feasibility in this module represents fast and effective method in nuclear safety analysis.
动态概率安全评估(D-PSA)在使用高保真的热液代码模拟复杂事故场景时面临着巨大的计算挑战,需要创新的方法来平衡准确性和效率。本研究引入了一个机器学习增强的动态事件树(DET)模块,通过整合深度学习和支持向量机(SVM)来改进风险分析的量化。该模块利用基于fortran的子模块自动生成和并行执行热水力模型输入文件,实现多个动态场景的并发模拟,而随机抽样策略确保以最小的训练数据需求全面覆盖故障序列。本研究以布什尔核电站的一次电站停电为起始事件,作为热工水力模型基准。在此基础上,对柴油发电机组在不同时间步长的动态回收效果进行了评价。因此,SBO演变为场外失电(LOOP)事故。提出的模块调查LOOP事故,以进行BNPP的综合风险分析。结果表明,所开发的机器学习架构取得了良好的预测性能,在分类堆芯损伤状态和减少场景评估时间方面准确率至少超过97%。该模块的高分辨率动态建模与计算可行性相结合,代表了核安全分析快速有效的方法。
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引用次数: 0
Exploration of grid types for radiation field representing and path-finding 辐射场表示和寻路网格类型的探索
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-12 DOI: 10.1016/j.pnucene.2026.106251
Jiamei Tang , Xiaodan Li , Mengkun Li , Li Liu , Mengxiao Wang
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引用次数: 0
Innovative design and safety evaluation of the decay heat removal system for the China fast reactor 中国快堆消热系统的创新设计与安全性评价
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-10 DOI: 10.1016/j.pnucene.2026.106238
Zhiwen Dai , Donghui Zhang , Yuting Yang , Zhiwei Zhou , Chao Lin , Xiuli Xue , Xintai Yu , Dalin Zhang , Xiyong Chen , Shangang Cao , Songping Wang , Chengwen Xing , Shuiwen Jiang
Pool-type sodium-cooled fast reactors (SFR) have become one of the main selections of Generation-IV reactors due to large thermal inertia and inherent safety, which solve the future shortage of natural uranium and the disposal challenges of spent nuclear fuel (SNF). The decay heat removal system (DHRS) is one of the most important safety systems and must be highly reliable. This study illustrates the design and innovations of the DHRS on the China Fast Reactor. A thermal-hydraulic analysis was conducted using the system program (named ERAC) under station blackout (SBO) conditions, and key parameters of the natural circulation process were evaluated. China's fast reactor design is innovative in many respects, and its novel DHRS design ensures the reactor's safety during emergencies. The analysis results show that the DHRS system operates effectively and that the calculations align with the design goals. Under natural circulation, the peak temperature reached approximately 592 °C at 1000 s. As natural circulation progressed, the core outlet temperature gradually decreased; by 5000 s, the average core fuel outlet temperature was 574 °C. The design of the core throttling component meets the requirements and can provide sufficient natural circulation. This study could provide a valuable reference for the design of SFRs.
池式钠冷快堆(SFR)由于热惯量大、固有安全性好,已成为第四代反应堆的主要选择之一,解决了未来天然铀短缺和乏核燃料(SNF)处理的难题。衰变排热系统(DHRS)是最重要的安全系统之一,必须具有高可靠性。本研究说明了中国快堆的DHRS的设计和创新。利用系统程序(ERAC)进行了电站停电条件下的热液分析,并对自然循环过程的关键参数进行了评价。中国的快堆设计在许多方面具有创新性,其新颖的DHRS设计确保了反应堆在紧急情况下的安全性。分析结果表明,该系统运行有效,计算结果符合设计目标。在自然循环条件下,1000 s温度峰值约为592℃。随着自然循环的进行,岩心出口温度逐渐降低;到5000s时,堆芯燃料出口平均温度为574℃。核心节流元件的设计满足要求,并能提供充分的自然循环。本研究可为SFRs的设计提供有价值的参考。
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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 Pub 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
Void fraction migration in developing bubbly flow through sudden cross-section contraction and expansion in square channels 方形通道中突然横截面收缩和膨胀的气泡流动中空隙率的迁移
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-09 DOI: 10.1016/j.pnucene.2025.106214
Ayodeji A. Ala , Lorlornyo Abusah , Shouxu Qiao , Bin Ye , Sichao Tan
Precise prediction of the flow pattern and proper estimation of the void fraction distribution are needed to design and optimize multiphase reactors and bubble columns. This research presents the cross-sectional and averaged void fraction distributions of adiabatic air-water mixtures across square-to-square channel cross-section expansion and contraction, with gas superficial velocity (jg) ranging from 0.003 m/s to 0.0054 m/s and water superficial velocity (jl) from 0.08 m/s to 0.36 m/s. A high-speed imaging system captures the bubble information, while key liquid phase flow features are extracted using the particle image velocimetry technique. Changing jg and jl have different effects on the bubble size distributions. In addition to the slowing flow after the expansion plane, the higher void fraction near the walls before the expansion influences the void fraction distribution in the expansion area. Void fraction distribution in channels with contraction displays a wall and core peaking profile. Conversely, a reduction in the contraction ratio in the channel increases the wall-peaking void fraction compared to the core. The decreasing averaged void fraction across the contraction transitions from linear to polynomial as the contraction in the channel increases. A code was developed to estimate the void fraction distribution before the contraction plane in a square channel, and the output was compared with available data.
在多相反应器和气泡塔的设计和优化中,需要精确的流态预测和合理的空隙率分布估计。研究结果表明,在气表速度(jg)为0.003 ~ 0.0054 m/s,水表速度(jl)为0.08 ~ 0.36 m/s的条件下,绝热空气-水混合物的横截面和平均空隙率分布。高速成像系统捕获气泡信息,同时利用粒子图像测速技术提取关键的液相流动特征。改变jg和jl对气泡尺寸分布有不同的影响。除了膨胀面后流动变缓外,膨胀前壁面附近较高的空隙率影响了膨胀区的空隙率分布。孔隙率在收缩通道中的分布表现为壁面和岩心峰值。相反,与岩心相比,通道中收缩比的减小增加了壁面峰值空隙率。随着通道收缩的增加,收缩过程中减少的平均空隙率从线性转变为多项式。开发了估算方形通道收缩面前空隙率分布的程序,并将输出结果与现有数据进行了比较。
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引用次数: 0
Development of a risk profile for accident sequences considering release-to-impact correlation 考虑释放-影响相关性的事故序列风险概况的发展
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-09 DOI: 10.1016/j.pnucene.2026.106242
Yunho Kim, Jaehyun Cho
Following the Three Mile Island accident, risk has become a fundamental concept in evaluating nuclear power plant (NPP) safety. Probabilistic Safety Assessment (PSA) is widely used to quantify risk by integrating accident frequencies and radiological consequences. However, the comprehensive PSA process, particularly level 3 consequence assessment using codes such as MACCS, demands significant computational resources and is not well-suited for rapid evaluations during design optimization. This challenge is intensified by recent Risk-Informed and Performance-Based (RIPB) regulatory frameworks, which require comprehensive risk quantification explicitly accounting for accident consequences across diverse accident sequences. To address this limitation, this study proposes a computationally efficient methodology based on a correlation model linking source-term releases to resulting radiological consequences. This approach enables the development of intuitive risk profiles that can enhance the practical utilization of risk information. As a case study, the proposed framework was applied to core damage accident sequences identified in the OPR-1000 level 1 PSA model, focusing on two initiating events: 1) loss of feedwater (LOFW), and 2) small break loss of coolant accident (SLOCA). Frequency-Consequence (F-C) curves were developed to compare the relative risks of the two events and evaluate the effects of severe-accident mitigation measures. The results demonstrate the initial applicability of the proposed methodology and indicate its potential to support more efficient risk-informed safety evaluations and decision-making.
在三里岛核事故之后,风险已经成为评价核电站安全的一个基本概念。概率安全评估(PSA)被广泛用于通过综合事故频率和辐射后果来量化风险。然而,综合PSA过程,特别是使用MACCS等代码进行的3级后果评估,需要大量的计算资源,并且不适合在设计优化期间进行快速评估。最近的基于风险和绩效(RIPB)的监管框架加剧了这一挑战,该框架要求对不同事故序列的事故后果进行全面的风险量化。为了解决这一限制,本研究提出了一种基于关联模型的计算效率方法,该模型将源项释放与由此产生的辐射后果联系起来。这种方法能够开发直观的风险概况,从而增强风险信息的实际利用。作为案例研究,提出的框架应用于OPR-1000 1级PSA模型中确定的堆芯损坏事故序列,重点关注两个初始事件:1)给水损失(LOFW)和2)冷却剂小破裂损失事故(SLOCA)。绘制了频率-后果(F-C)曲线,以比较两种事件的相对风险,并评估严重事故缓解措施的效果。结果证明了所提出方法的初步适用性,并表明其支持更有效的风险知情安全评估和决策的潜力。
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引用次数: 0
Cybersecurity risk assessment of nuclear SVDU based on attack tree 基于攻击树的核SVDU网络安全风险评估
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-09 DOI: 10.1016/j.pnucene.2026.106241
Yanqun Wu , Junfeng Wang , Quan Ma , Mingxing Liu , Meiyuan Chen
The nuclear safety video display unit (SVDU) is a critical human-machine interface in the nuclear reactor protection system, and its cybersecurity is increasingly challenged by the sector's full digitalization. This paper proposes a novel cybersecurity risk assessment method integrating an attack tree model with the STRIDE threat framework to evaluate the risk status of the nuclear SVDU. The approach innovatively incorporates the latest CVSS4.0 standard to define security attributes (attack cost, technical difficulty, discovery difficulty, and impact) for leaf nodes, which are quantified through hierarchical scoring. Crucially, an objective entropy weighting method is employed to calculate the weights of these attributes, effectively eliminating the subjective bias inherent in traditional expert-dependent methods like FAHP. The risk probability is then propagated from leaf to root nodes based on their dependency relationships. Comparative analysis demonstrates that the proposed method offers a more sensitive and discriminating assessment, identifying critical threats that subjective approaches might overlook and providing a robust, data-driven basis for prioritizing security protection strategies.
核安全视频显示单元(SVDU)是核反应堆防护系统中关键的人机界面,随着该领域的全面数字化,其网络安全日益受到挑战。本文提出了一种将攻击树模型与STRIDE威胁框架相结合的新型网络安全风险评估方法,以评估核SVDU的风险状态。该方法创新性地结合了最新的CVSS4.0标准来定义叶节点的安全属性(攻击成本、技术难度、发现难度和影响),并通过分层评分对其进行量化。最关键的是,采用客观熵加权法计算这些属性的权重,有效消除了传统专家依赖方法(如FAHP)固有的主观偏差。然后根据它们的依赖关系将风险概率从叶节点传播到根节点。对比分析表明,所提出的方法提供了一个更敏感和有区别的评估,识别主观方法可能忽略的关键威胁,并为优先考虑安全保护策略提供了一个健壮的、数据驱动的基础。
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引用次数: 0
Uncertainty quantification for a vessel fluence calculation: A PWR case study 容器流量计算的不确定性量化:压水堆案例研究
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-07 DOI: 10.1016/j.pnucene.2025.106234
D. Timpano , A. Vasiliev , D. Rochman , M. Hursin
The European project Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT (EVEREST) was launched in 2024 with the goal to employ advanced multiphysics tools to contribute to Long Term Operation (LTO) of nuclear power plants (NPP). The improvement of the reactor pressure vessel (RPV) fluence calculations is one of the overarching aims of this research endeavor, alongside a quantification of the modeling biases and uncertainties related to the standard computational methods. In the framework of the EVEREST project, EPFL has set up a new methodology for vessel fluence calculation based on Polaris/PARCS/Serpent and applied it to the Turkey Point 3 Pressurized Water Reactor (PWR), leveraging on public core design and operating data. This paper aims at presenting the results of the sensitivity analysis and uncertainty quantification performed for this application case. A novel integrated workflow to propagate uncertainties in the whole computational chain of vessel fluence calculations is described. The analysis covers core follow modeling, source preparation and shielding calculations, tackling the uncertainty due to material densities, geometry and nuclear data. Both standard Sandwich Formula (SF) and Total Monte-Carlo (TMC) techniques have been used for this scope. The UQ performed in this study revealed an analytical uncertainty associated with the fast flux at the thermal shield and the reactor pressure vessel location between 8 and 10%. The main contributors are the uncertainties on nuclear data and manufacturing tolerances in shielding calculations. This work contributes to the implementation of the Best Estimate Plus Uncertainty (BEPU) approach for assessing radiation damage in nuclear reactor structural materials.
欧洲项目验证和增强反应堆压力容器流量评估实验(EVEREST)于2024年启动,目标是采用先进的多物理场工具,为核电站(NPP)的长期运行(LTO)做出贡献。改进反应堆压力容器(RPV)流量计算是本研究的首要目标之一,同时量化与标准计算方法相关的建模偏差和不确定性。在EVEREST项目的框架内,EPFL建立了一种基于Polaris/PARCS/Serpent的容器流量计算新方法,并利用公开的堆芯设计和运行数据将其应用于土耳其3号压水堆(PWR)。本文旨在介绍对该应用案例进行敏感性分析和不确定度量化的结果。提出了一种新的集成工作流,将不确定性传播到整个容器流量计算链中。分析包括核心跟随建模,源准备和屏蔽计算,解决由于材料密度,几何形状和核数据的不确定性。标准三明治公式(SF)和全蒙特卡罗(TMC)技术已用于此范围。在本研究中进行的UQ显示,与热屏蔽处的快速通量和反应堆压力容器位置相关的分析不确定度在8%到10%之间。主要原因是核数据的不确定性和屏蔽计算中的制造公差。这项工作有助于实现核反应堆结构材料辐射损伤评估的最佳估计加不确定性(BEPU)方法。
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引用次数: 0
The α-radiolysis behavior and mechanism of tri-isoamyl phosphate (TiAP): New insights from 4He2+ beam irradiation experiments and theoretical calculations 磷酸三异戊酯(TiAP) α-辐射分解行为和机理:来自4He2+辐照实验和理论计算的新见解
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-06 DOI: 10.1016/j.pnucene.2025.106233
Shilong Shi , Yaoshuang Wan , Guofeng Qu , Runyu Zhang , Jiajian Song , Xuanhao Huang , Jijun Yang , Zhihui Li , Tu Lan , Songdong Ding , Yuanyou Yang , Jiali Liao , Wen Feng , Jing Peng , Ning Liu
Understanding the α-radiolysis of extractants is crucial for their applications in nuclear fuel reprocessing. However, current methodologies face challenges in conducting α-irradiation experiments and elucidating the underlying mechanisms. Herein, a 4–7 MeV 4He2+ beam provided by a CS-30 cyclotron was employed as a fast, convenient, and versatile α-irradiation source to simulate actual radiative scenarios in nuclear fuel reprocessing, which was firstly employed to investigate the α-radiolysis of tri-iso-amyl phosphate (TiAP) in alkane diluent, an alternative extractant in Plutonium Uranium Recovery by Extraction. The dominant radiolysis products of TiAP including hydrogen, methane and di-iso-amyl phosphate (DiAP) were qualitatively and quantitatively determined by GC and HR-MS. Ehrenfest dynamic simulations and DFT calculations revealed the ionization of TiAP induced by electronic stopping is the dominant process in its α-radiolysis process, as confirmed by experimental results. The C‒O bond cleavage in TiAP, leading to the formation of DiAP, is attributed to the decomposition of TiAP+ and TiAP∗, as well as reactions between TiAP and ·OH or secondary electrons. Finally, the radiolysis mechanism of TiAP in alkane diluent was proposed based on the analysis of radiolytic products and multi-time-scale theoretical calculations. This study paves the way for advancing research on the α-radiolysis of extractants for spent fuel reprocessing.
了解萃取剂的α-辐射分解对其在核燃料后处理中的应用至关重要。然而,目前的方法在进行α-辐照实验和阐明潜在机制方面面临挑战。本文采用CS-30回旋加速器提供的4-7 MeV 4He2+束流作为快速、方便、通用的α-辐照源,模拟了核燃料后处理过程中的实际辐射情景,并首次应用该辐照源研究了三异戊基磷酸(TiAP)在提取法回收钚铀的替代萃取剂烷烃稀释剂中的α-辐射分解。采用气相色谱(GC)和质谱(HR-MS)对TiAP的主要辐射分解产物氢、甲烷和磷酸二异戊酯(DiAP)进行了定性和定量分析。Ehrenfest动力学模拟和DFT计算表明,TiAP α-辐射分解过程中电子停止引起的电离是主要过程,实验结果也证实了这一点。TiAP中C-O键的断裂导致了DiAP的形成,这归因于TiAP+和TiAP *的分解,以及TiAP与·OH或二次电子之间的反应。最后,通过辐射分解产物分析和多时间尺度理论计算,提出了TiAP在烷烃稀释剂中的辐射分解机理。本研究为推进乏燃料后处理萃取剂α-辐射分解的研究奠定了基础。
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引用次数: 0
Fuzzy adaptive sliding mode control of U-tube steam generator water level based on a nonlinear dynamic model 基于非线性动态模型的u型管蒸汽发生器水位模糊自适应滑模控制
IF 3.2 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-06 DOI: 10.1016/j.pnucene.2025.106235
Ze Zhu , Xiaojie Guo , Zhiwu Ke , Kelong Zhang , Pengfei Wang
The false water level caused by the so-called “swell and shrink”, and the measurement errors, are the two major challenges in the water level control of U-tube steam generators (UTSGs). This paper proposes a fuzzy adaptive sliding mode controller (FASMC), which effectively mitigates the adverse impacts of false water levels, thus improving the water level control performance. Firstly, an ideal SMC (ISMC) was designed based on a detailed nonlinear UTSG model. Then, to reduce the adverse impacts of measurement errors, a water level estimator was constructed based on the backstepping and Lyapunov's second method, and an adaptive SMC (ASMC) was designed. Finally, the fuzzy algorithm was used to calibrate the hyperparameters in the ASMC online. The simulation results show that the established ISMC, ASMC, and FASMC outperform the PID controller under different operating conditions, with the integrated absolute errors (IAEs) of water level reduced by at least 5.68 %. And the FASMC has the best performance, with the IAE reduced by up to 74.8 %. This demonstrates the effectiveness and superiors of the proposed FASMC, which can provide a valuable reference for mitigating the adverse impacts of false water level in engineering practices and improving the UTSG water level control performance.
所谓“胀缩”引起的假水位和测量误差是u型管蒸汽发生器水位控制面临的两大挑战。本文提出了一种模糊自适应滑模控制器(FASMC),有效地减轻了假水位的不利影响,从而提高了水位控制性能。首先,基于详细的非线性UTSG模型,设计了理想SMC (ISMC)。然后,为了减小测量误差的不利影响,基于反推法和Lyapunov第二方法构造了水位估计器,并设计了自适应SMC (ASMC)。最后,利用模糊算法对ASMC超参数进行在线标定。仿真结果表明,所建立的ISMC、ASMC和FASMC在不同工况下均优于PID控制器,水位的综合绝对误差(iae)至少降低了5.68%。其中FASMC的性能最好,IAE降低了74.8%。验证了该方法的有效性和优越性,可为工程实践中减轻假水位的不利影响,提高UTSG水位控制性能提供有价值的参考。
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引用次数: 0
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Progress in Nuclear Energy
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