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

作者： Chao, Guo Jian, Deng Rong, Cai Yugao, Ma Lili, Liu Yuhao, Zhang Siyu, Lv Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Nuclear Science and Engineering School North China Electric Power University Changping District Beijing102206 China

ISBN: (纸本)9780791883785

According to spray cooling characteristics for failure of the cooling system in the spent fuel pool, we establish a model for analyzing spray cooling of a single spent fuel rod and develop a code named SFSC which is used to calculate the thermalhydraulic characteristics of film on the spent fuel rod under the condition of spray cooling. The results obtained from SFSC consist of film thickness, film temperature, film evaporation, fuel pellet temperature of spent fuel rod and cladding temperature of spent fuel rod at different height of spent fuel rod are shown in this paper. At the same time, we establish same model by RELAP and compare results calculated by RELAP with that calculated by SFSC. The good agreement of two results demonstrates that SFSC is effective. SFSC fills in the blanks of thermal-hydraulic calculation for the spent fuel rod under the condition of spray cooling. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Cooling

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Learning Pre-Grasp Manipulation of Flat Objects in Cluttered Environments using Sliding Primitives

Learning Pre-Grasp Manipulation of Flat Objects in Cluttered...

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IEEE International Conference on Robotics and Automation (ICRA)

作者： Jiaxi Wu Haoran Wu Shanlin Zhong Quqin Sun Yinlin Li State Key Laboratory of Turbulence and Complex Systems Intelligent Biomimetic Design Lab College of Engineering Peking University Beijing China Department of Automation University of Science and Technology of China Heifei China State Key Laboratory of Management and Control for Complex System Institute of Automation Chinese Academy of Science Beijing China Science and Technology on Thermal Energy and Power Laboratory Wuhan China

Flat objects with negligible thicknesses like books and disks are challenging to be grasped by the robot because of the width limit of the robot's gripper, especially when they are in cluttered environments. Pre-grasp manipulation is conducive to rearranging objects on the table and moving the flat objects to the table edge, making them graspable. In this paper, we formulate this task as Parameterized Action Markov Decision Process, and a novel method based on deep reinforcement learning is proposed to address this problem by introducing sliding primitives as actions. A weight-sharing policy network is utilized to predict the sliding primitive's parameters for each object, and a Q-network is adopted to select the acted object among all the candidates on the table. Meanwhile, via integrating a curriculum learning scheme, our method can be scaled to cluttered environments with more objects. In both simulation and real-world experiments, our method surpasses the existing methods and achieves pre-grasp manipulation with higher task success rates and fewer action steps. Without fine-tuning, it can be generalized to novel shapes and household objects with more than 85% success rates in the real world. Videos and supplementary materials are available at https://***/view/pre-grasp-sliding.

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Predicting Reverse Flow Characteristics in Equal-Height-Difference Passive Heat Removal system for Ocean Nuclear Power

SSRN

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SSRN 2022年

作者： Wang, Zhiwei He, Yanping Duan, Zhongdi Huang, Chao Li, Mingzhi Liu, Shiwen Hao, Chengming State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai200240 China Institute of Marine Equipment Shanghai Jiao Tong University Shanghai200240 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The equal-height-difference passive heat removal system (EHDPHRS) via natural circulation driving and utilizes ocean as infinite heat sink for heat dissipation, showing broad application prospect in ocean nuclear power field;but the heat removal pipe normally encounters the reverse flow of the cold seawater into the pipe, leading to operation instability and even condensation-induced water hammer. In this study, the integration of the volume of fluid model and the condensation model based on surface renewal theory is used to simulate the reverse flow characteristics in heat removal pipe under single-and two-phase natural circulation conditions. The interfacial heat and mass transfer are calculated by user-defined functions. The numerical results reveal that the buoyancy force ( [[EQUATION]] ) and condensation driving force ( [[EQUATION]] ) both accelerate the reverse flow, while the inertial force ( [[EQUATION]] ) restrains the reverse flow;the reverse flow length presents monotonous decrease with increasing flow velocity under single-phase conditions and parabolic-like trend for two-phase flow. A prediction correlation for reverse length is developed using dimensionless Reynolds number, Grashof number and vapor volume fraction, covering single- and two-phase conditions. The predicted reverse length by the correlation agrees with 94% of the results within deviations of [[EQUATION]] 15%. © 2022, The Authors. All rights reserved.

关键词： Forecasting

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Development of self-healing coating on fuel cladding for water-cooling nuclear reactors 14

Development of self-healing coating on fuel cladding for wat...

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14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Duan, Z.G. Xin, G.S. Zhou, Y. Li, Y.M. Abe, H. Yang, H.L. Kano, S. McGrady, J. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Nuclear Professional School School of Engineering University of Tokyo Tokai Ibaraki319-1188 Japan

A novel scheme for a bilayer coating with self-healing ability is proposed in this study. The candidate materials for the coatings and the potential self-healing reaction are assessed in high-temperature aqueous environments and high-temperature air. The pure Cr2O3 layer and the composite of Cr2O3 and MoO3 are the candidate materials for the outer layer and inner layer, respectively, due to their compatibility under normal condition. Fe2O3-MoO3 reactions exhibit a potential ability to heal the cracks because of a high reaction rate under normal condition. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Light water reactors

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Stratification and heat transfer of molten corium pool for in-vessel retention

Stratification and heat transfer of molten corium pool for i...

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

作者： Xiang, Qingan Deng, Jian Zhu, Dahuan Wu, Xiaoli Bi, Jinsheng Chen, Baowen Cai, Rong Qian, Libo YugaoMa Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Nuclear and radiation safety center Ministry of Environment Protection of P.R. China Beijing China

ISBN: (纸本)9780791883778

In-vessel retention (IVR) consists in cooling the molten corium contained in the lower head of reactor vessel by natural convection and reactor cavity flooding. The general approach which is used to study IVR problems is a "bounding" approach which consists in assuming a specified corium pool stratification in the vessel and then demonstrating that the vessel can cope with the resulting thermal and mechanical loads. Thermal loading on the vessel is controlled by the convective heat transfer inside the molten corium pool. Traditionally molten corium pool in the lower head was expected to stratify into two-layer with the dense oxide pool at the bottom and the light metal layer on the top. Based on the MASCA experiments, the increased density of the metal layer attributed to a transfer of uranium metal leads to inverse stratification with a heavy metal layer relocating below the oxide pool. This behavior can be explained by physicochemical interaction between the oxidic and metallic phases of the corium pool. Therefore, a methodology which couples physicochemical effects and thermal hydraulics has been developed to address the IVR issue. The main purpose of this paper is to present this methodology and its application for calculate stratification probability of two-layer and three-layer configuration, analyze the safety margin of IVR in two-layer and three-layer configuration, and evaluate the lower head heat thermal failure probability. Copyright © 2020 ASME

关键词： Heat flux

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The stress indices study for the pipe with trunnion welding attachment 27

The stress indices study for the pipe with trunnion welding ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Lu, Xifeng Bai, Xiaoming Zheng, Liangang Zhang, Rui Yuan, Feng Yuan, Yanli Huang, Qian Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888983051

In the nuclear power plant, the trunnion welding attachment is widely used to connect the pipe and support. According to ASME Code Cases the stress caused by the welded attachment is considered as the additional stress of the pipe. However, the calculation of the additional stress is very complex and time-consuming. A method based on the concept of stress indices is proposed in present work. To investigate the influence of size and constraint and the stress indices, the finite element analysis for pipe with different sizes trunnion welding attachment under different constraint forms are done. The stress indices are calculated for each case, respectively. Furthermore, the fitted expressions for stress indices are proposed for general case. To verify this method, an analysis according to ASME Code Case for piping with trunnion is done. The comparison of the result between present method and the ASME Code Case method show slight different. Therefore, the stress indices method for the stress analysis of pipe with trunnion welding attachment can be an alternative method for ASME Code Cases. The advantage of present method in that it simplified the analysis process by introduces the stress indices. This method is effective and practical during the design and analysis of piping system in the nuclear power plant. Copyright © 2019 by JSME

关键词： Welding

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Uncertainty quantification research of reactor system under seismic load 27

Uncertainty quantification research of reactor system under ...

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27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Huang, Qian Xiong, Furui Bai, Xiaoming Yu, Xiaofei Zheng, Liangang Lu, Xifeng Song, Haiyang Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888983051

In the dynamic analysis of reactor system, the key mechanical input parameters are uncertain due to the design change, manufacturing, installation and calculation deviation, which affects the dynamic response, load distribution and final mechanical evaluation results. In order to quantify the influence of the uncertainty on seismic responses and load distribution, the uncertainty quantification of key structural input parameters of reactor system under seismic load is studied. Firstly, according to the fundamental characters of key input parameters (contact stiffness and gap among various assemblies), maximum entropy principle is adopted to construct the probability density function(PDF). Then, Markov Chain Monte Carlo technique is applied to sample according to the PDF, and the input and output data pool is constructed via finite element transient computations. According to the results, probabilistic distributions of dynamic responses caused by uncertainty from inputs parameters are evaluated and the reliability of nominal model is investigated. The studies show that the probabilistic distribution of reactor system dynamic response differs greatly with different output mechanical parameters, degrees of freedom and frequency bands. The uncertainty quantification method and process for reactor system are proposed in this paper. This method can be extended to other equipments easily. Therefore, it is an effective tools for the reliability of dynamic response. Copyright © 2018-2019 by JSME

关键词： Markov processes

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Correction to:Study on the resource utilization of high fluorine-containing organic waste through fluidized incineration

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Waste Disposal and Sustainable Energy 2023年第2期5卷 245-245页

作者： Wenhan Li Zengyi Ma Jianhua Yan Qianming Huang Xingjian Wen Zian Zhai Bochen Huang Shuang Wang Yongqiang Chen Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu610213China State Key Laboratory of Clean Energy Utilization Institute for Thermal Power EngineeringZhejiang UniversityHangzhou310027China Zhejiang Xianchuang Energy&Technology Co.Ltd Jinhua321000China

Correction to:Waste Disposal&Sustainable Energy(2022)4:117-129 https://***/10.1007/s42768-022-00101-7 The section Confict of interest'has been amended:'Jian-hua Yan is the Editor in-Chief of Waste Disposal&Sustain-able Energy.'The revised'Confict of interest'is as follows:Jianhua Yan is the Editor-in-Chief of Waste Disposal&Sustainable *** behalf of all authors,the corresponding author states that there is no conflict of interest.

关键词： Sustainable fluorine behalf

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Effect of thermal-hydraulic parameters and pH on the corrosion product deposition in steam generator based on CFD

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Annals of Nuclear Energy 2025年 223卷

作者： Yunchen Lai Haiyan Xie Shiyu Tan Ke Wang Dongyun Li Xuesong Li Yang Gao Hongguo Hou Nan Chao Caishan Jiao Yu Zhou Fundamental Science on Nuclear Safety and Simulation Technology Laboratory College of Nuclear Science and Technology Harbin Engineering University Harbin 150001 Heilongjiang Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610041 Sichuan Province China

The deposition of corrosion products in the primary circuit of a pressurized water reactors is known to have an adverse impact on reactor system’s performance, which can cause localized corrosion. They can also impede the flow of coolant, or in severe cases block the tube channels. Besides, the deposits have a lower thermal conductivity than the structural materials, which may increase the thermal resistance, thus reducing heat transfer efficiency. Despite the presence of a considerable amount of corrosion products in steam generators, research on the deposition of corrosion products within is still insufficient. A comprehensive 3-D model of thermal–hydraulic and chemical processes was used to investigate the factors affecting the deposition of corrosion products in a steam generator. For the purpose of determine the deposition distribution along the heat transfer tube, the tube was separated into 11 main nodes based on the equal temperature differences according to the thermal–hydraulic data calculated. The simulation results indicate that the deposition thickness is sensitive to coolant temperature, flow velocity and pH value, and it was shown that higher temperature(from 560 K to 600 K decreasing 9.31 %), pH value (from 6.5 to 8.0 decreasing 9.92 %) and lower flow velocity (from 5 m/s to 20 m/s increasing 8.13 %)was effective for reducing the corrosion product deposition due to the physicochemical properties of the coolant and the turbulent flow, while working pressure has less influence on it(from 6 MPa to 15 MPa increasing only 1.03 %).

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Study in pool of research cooling reactor strategy at for accident pressure situation vessel

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Kerntechnik 2020年第2期85卷 88-97页

作者： Zhang, D. Ran, X. Zhang, Y. Zhang, H. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

Pressure vessel in pool layout is usually used in high power and middle pressure research reactor, Flow inversion and residual heat removal is important problem. The heat removal strategy at accident situation for a pressure vessel in pool layout research reactor was studied in the paper, many cases was researched for the ability of heat removal and safety of core. © Carl Hanser Verlag, München.

关键词： Research reactors

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