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arXiv 2025年

作者： Zhang, Ruizhi Zhu, Shengfeng Wang, Kan She, Ding Argaud, Jean-Philippe Bouriquet, Bertrand Li, Qing Gong, Helin School of Mathematical Sciences Shanghai Key Laboratory of Pure Mathematics and Mathematical Practice East China Normal University Shanghai200241 China Department of Engineering Physics Tsinghua University Beijing100084 China Électricité de France R&D 7 boulevard Gaspard Monge Palaiseau91120 France Électricité de France DQI 2 rue Ampère Saint-Denis93206 France Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Paris Elite Institute of Technology Shanghai Jiao Tong University Shanghai200240 China

reactor physics is the study of neutron properties, focusing on using models to examine the interactions between neutrons and materials in nuclear reactors. Artificial intelligence (AI) has made significant contributions to reactor physics, e.g., in operational simulations, safety design, real-time monitoring, core management and maintenance. This paper presents a comprehensive review of AI approaches in reactor physics, especially considering the category of Machine Learning (ML), with the aim of describing the application scenarios, frontier topics, unsolved challenges and future research directions. From equation solving and state parameter prediction to nuclear industry applications, this paper provides a step-by-step overview of ML methods applied to steady-state, transient and combustion problems. Most literature works achieve industry-demanded models by enhancing the efficiency of deterministic methods or correcting uncertainty methods, which leads to successful applications. However, research on ML methods in reactor physics is somewhat fragmented, and the ability to generalize models needs to be strengthened. Progress is still possible, especially in addressing theoretical challenges and enhancing industrial applications such as building surrogate models and digital twins. Copyright © 2025, The Authors. All rights reserved.

关键词： Nuclear industry

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Experimental and numerical investigation on the flow and heat transfer characteristics of lead-bismuth eutectic through a helical single rod and four-rod bundle

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International Journal of Heat and Mass Transfer 2025年 251卷

作者： Dianqiang Jiang Yiwen Chen Dalin Zhang Kunfeng Li Jian Deng Qi Lu Xiaoli Wu Ting Zhang Wenxi Tian State Key Laboratory of Multiphase Flow in Power Engineering Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University Xi’an 710049 China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu 610213 China

The flow and heat transfer performance of nuclear fuel are critical factors influencing the safety and economic efficiency of nuclear reactor systems. This study investigates a novel helical fuel design for lead-bismuth eutectic-cooled reactors. Experiments are performed in three different test sections to evaluate flow resistance, convective heat transfer, transverse mixing, and flow distribution. key experimental parameters, including temperature, pressure drop, and volumetric flow rate, are measured. The results demonstrate that, at a constant Reynolds number ( Re ) and constant Péclet number ( Pe ), the inlet temperature has negligible influence on both the resistance coefficient and Nusselt number ( Nu ). The flow resistance correlation is found to be consistent across fuel rods similar to helical cruciform geometry. Besides, transverse mixing experiments in the four-rod test section demonstrate that transverse mixing efficiency across subchannels is influenced by experimental parameters. Moreover, flow distribution results indicate that the flow distribution ratio (average = 1.89) between inside and outside channels remains independent of Re in turbulent regions. These experiments provide a comprehensive dataset for validating numerical models. Based on these data, computational fluid dynamics (CFD) simulations are performed. The predictions using the Realizable two-layer k − ε turbulence model show excellent agreement with experimental data.

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Q-Learning for Finite-Horizon H∞ Tracking Control of Unknown Linear Discrete-Time systems

Q-Learning for Finite-Horizon H∞ Tracking Control of Unknow...

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Chinese Control Conference (CCC)

作者： Mingxiang Liu Qianqian Cai Wei Meng Dandan Li Minyue Fu Zhimin Han School of Automation Guangdong University of Technology Guangzhou China Guangdong Province Key Laboratory of Intelligent Decision and Cooperative Control Guangzhou China School of System Design and Intelligent Manufacturing Southern University of Science and Technology Shenzhen China School of Automation Hangzhou Dianzi University Hangzhou China

In this paper, we present a Q-learning algorithm for finite-horizon H∞ tracking control of unknown linear discrete-time systems. Finite horizon control is challenging due to its correspondence with a time-varying Riccati equation. The proposed algorithm can determine the solution of the time-varying Riccati equation with unknown system dynamics. We first establish the time-varying Q-function, then propose a corresponding Q-learning algorithm and prove its convergence. Finally, the proposed algorithm is proved to be effective through numerical examples.

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Analysis of Thermal-Hydraulic Best-Estimate Program Against the Downcomer Countercurrent Flow Limitation

SSRN

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SSRN 2022年

作者： Li, Xiang Sun, Wan Ding, Shuhua Huang, Tao Ma, Zaiyong Zhang, Luteng Zhu, Longxiang Pan, Liangming Key Laboratory of Low-Grade Energy Utilization Technologies & Systems Chongqing University Ministry of Education Chongqing400044 China Department of Nuclear Engineering and Technology Chongqing University Chongqing400044 China Key laboratory of nuclear reactor system design technology Nuclear Power Institute of China Chengdu610044 China

The relevance of the countercurrent flow limitation phenomenon is known in nuclear (emergency core cooling and accident situations) industries, such as a Loss of Coolant Accident (LOCA). The occurrence of countercurrent flow limitation (CCFL), which can exist in a host of places around the nuclear reactor system, will prevent the emergency cooling water, which is necessary to cool the core of successfully flowing into the core from the downcomer of Pressurized Water reactor (PWR) in the LOCA. Accurate prediction of the downcomer CCFL phenomenon is significant for the design and analysis of reactor safety. Among the experiments at Upper Plenum Test Facility (UPTF), various tests, test 6 and test 7 are studied in this research. The conservatively evaluated using the best-estimate program RELAP5/MOD3.3 to predict the CCFL phenomenon at the downcomer. Above all, the methodology of node partition for the channels with cross-connections to represent the circumferential downcomer annulus was also studied. In particular, the nodes along circumferential direction had a vital effect on the prediction of the steam velocity profile of downcomers. The influence of the CCFL model, including the most widely used Wallis-type and Kutateladze-type correlations, and the cooling limited liquid-downflow behavior were analyzed. In addition, the pressure on the downcomer and steam flow rate of the break were monitored, and the predicted results were also analyzed. © 2022, The Authors. All rights reserved.

关键词： Loss of coolant accidents

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Interfacial area transport model for bubbly flow system in vertical rod bundle with spacer grids

Interfacial area transport model for bubbly flow system in v...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Liu, Hang Lai, Jianyong Li, Yi Zhang, Yulong Liu, Minghao Pan, Liangming Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Nuclear Engineering Chongqing University Chongqing400044 China

ISBN: (纸本)9780791883785

In relation to the modeling of the one-dimensional interfacial area transport equation for an adiabatic bubbly flow in a vertical rod bundle, some existing models of sink and source terms were reviewed and evaluated. Based on the reviewed interaction mechanisms of bubbles and turbulent eddies, a new interfacial area transport model has been proposed. Two important impacts on bubble interaction have been taken into account in the new model: The effects of spacer grids with mixing vanes and the impacts of geometry structure. The spacer grids breakup large bubbles into small bubbles resulting in enhanced bubble random collision at the downstream of the spacer grids. Void transport is the main contribution between spacer grids. The new interfacial area transport model has been evaluated against the obtained experimental data in 17 bubbly flow conditions. The results indicate that the new model can predict the interfacial area concentration with the relative error of 19.9%. It is recognized that the proposed model is promising for predicting the interfacial area concentration for a bubbly flow in a rod bundle. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear engineering

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Multiple choices of reactor core nuclear design for ACP100's application in different scenarios

Multiple choices of reactor core nuclear design for ACP100's...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Wang, Liangzi Ju, Haitao Li, Qing Qin, Dong Wang, Lianjie Yu, Yingrui Ning, Zhonghao Wang, Chenlin Guo, Rui Wang, Shuai Zhang, Bin Xiang, Hongzhi Lou, Lei Sun, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China NO. 328 Section 1 Changshun Ave Shuangliu District Sichuan Chengdu China

ISBN: (纸本)9781713827245

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology;it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design;thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for da ily boron adjustment and relative waste storage;choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate t he core, through lar

关键词： Boron

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RMC/ANSYS Multi-physics coupling solutions for heat pipe cooled reactors analyses

RMC/ANSYS Multi-physics coupling solutions for heat pipe coo...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Ma, Yugao Liu, Minyun Chen, Erhui Xie, Biheng Chai, Xiaoming Huang, Shanfang Wang, Kan Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781713827245

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion. © The Authors, published by EDP sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://***/licenses/by/4.0/).

关键词： Finite element method

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Noncontact Doppler Radar-based Heart Rate Detection on the SVD and ANC

Noncontact Doppler Radar-based Heart Rate Detection on the S...

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2022 IEEE International Conference on Bioinformatics and Biomedicine, BIBM 2022

作者： Shi, Qiuxia Hu, Bin Tian, Fuze Zhao, Qinglin Gansu Provincial Key Laboratory of Wearable Computing School of Information Science and Engineering Lanzhou University China School of Computer and Communication Lanzhou University of Technology China Key Laboratory of Special Functional Materials and Structural Design Lanzhou University Ministry of Education Lanzhou China CAS Center for Excellence in Brain Science and Intelligence Technology Shanghai Institutes for Biological Sciences Chinese Academy of Sciences China Joint Research Center for Cognitive Neurosensor Technology of Lanzhou University Institute of Semiconductors Chinese Academy of Sciences China Engineering Research Center of Open Source Software and Real-Time System Lanzhou University Ministry of Education China

ISBN: (纸本)9781665468190

In the Doppler biological radar-based applications of noncontact measurement of vital signs, effectively extracting heartbeat information from weak thoracic mechanical motion is an important problem to be solved. This study is aimed at extracting heartbeat signal via the technology combined with Short Time Fourier Transform (STFT), Singular Value Decomposition (SVD) and Adaptive Noise Canceller (ANC) from radar recording. The simulated data and the data collected by Doppler radar biosensor realized in laboratory are employed to validate the proposed method. The results show that the proposed method has the ability of detection for the heart rate and heart rate variability indexes in rest state, it has certain advantages in time-consuming and detection accuracy. Therefore, the current method provides another way to process vital sign signals recorded by Doppler radar. © 2022 IEEE.

关键词： Singular value decomposition

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Parametric Structural Optimization of 2D Complex Shape Based on Isogeometric Analysis

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Computer Modeling in Engineering & sciences 2020年第7期124卷 203-225页

作者： Long Chen Li Xu Kai Wang Baotong Li Jun Hong School of Mechanical Engineering University of Shanghai for Science and TechnologyShanghai200093China Key Laboratory of Education Ministry for Modern Design&Rotor-Bearing System Xi’an Jiaotong UniversityXi’an710049China

The geometric model and the analysis model can be unified together through the isogeometric analysis method,which has potential to achieve seamless integration of CAD and *** design is a mainstream and successful method in CAD *** method is not continued in simulation and optimization stage because of the model conversion in conventional optimization method based on the finite element *** integration of the parametric modeling and the structural optimization by using isogeometric analysis is a natural and interesting *** paper proposed a method to realize a structural optimization of parametric complex shapes by using isogeometric *** the given feature curves and the constraints,a feature frame model is *** on the feature frame model,a parametric representation of complex shape is *** adding some auxiliary curves,the feature frame model is divided into many box-like patches in three dimension or four-sided patches in two *** patches are built into parametric patches by using volume interpolation methods such as Coons *** on the parametric patches,isogeometic analysis is ***,the relationships are constructed among the size parameters,the control points and the physical performance *** the sensitivity matrix could be derived based on the *** size optimization is carried out in the first stage by taking the size parameters as *** on the result of size optimization,shape optimization with the constraints of stress is carried out in the second stage by taking the control points as *** planar complex shapes are taken as example to verify our *** results verify that the parametric modeling and structural optimization can be united together without model *** from this,the optimization design can be executed as a dark box operation without considering the concrete modeling and analysis by input of the sizes,constra

关键词： Isogeometric analysis parametric design structural optimization

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Addition of Niobium in Fe-13cr-4.5al-2mo Alloy Used as Atf Cladding: Effect on High Temperature Water Corrosion and In-Situ Electrochemistry

SSRN

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SSRN 2022年

作者： Liao, Jingjing Wang, Hui Wu, Jun Zhang, Wei Xu, Fei Sun, Hongying An, Xuguang Qiu, Shaoyu Chengdu University China Science and Technology on Reactor Fuel and Materials Laboratory Nuclear Power Institute of China Sichuan Chengdu610213 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu610213 China School of Mechanical Engineering Anyang Institute of Technology Henan Anyang455002 China Interdisciplinary Materials Research Center Institute for Advanced Study Chengdu University Sichuan Chengdu610106 China

Niobium addition in candidate accident tolerant fuel (ATF) cladding FeCrAl alloy has received much attention, but its effect on high temperature water corrosion behavior has not been revealed. Four alloys with different Nb content (0wt%, 0.5wt%, 1.0wt% and 1.5wt%) based on Fe-13Cr-4.5Al-2Mo model alloy were designed, and the corrosion behavior was investigated by in-situ electrochemistry and microscopy. The in-situ electrochemistry analyses were carried out in 360 。C/19.0MPa lithiated and borated water connected with a loop system to monitor the water chemistry. Detailed studies on the oxide film formed after exposure for 85 days were conducted by focused ion bombarding and transmission electron microscopy. FeCrAl alloys have satisfactory corrosion resistance because of small corrosion currents and thin oxide films. Only monolayer multicrystal spinel oxide forms on alloy surface. Slight Nb addition (0.5% and 1.0wt%) benefits the corrosion resistance due to relatively large impedance of oxide and delayed oxidation of Fe2 Nb particles. Porosities with Fe depletion are spotted in oxide and especially severe in 1.5wt%-Nb alloy. Cathodic Tafel kinetics and corrosion mechanisms of FeCrAl are further discussed. © 2022, The Authors. All rights reserved.

关键词： Oxide films

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