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

作者： Ma, Yugao Liu, Minyun Han, Wenbin Xie, Biheng Chai, Xiaoming Huang, Shanfang Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Space fission power systems can enable ambitious solarsystem and deep-space science missions. The heat pipe cooled reactor is one of the most potential candidates for near-term space power supply, featured with safety, simplicity, reliability, and modularity. Heat pipe cooled reactors are solid-state and high temperature (up to 1500 K) reactors, where the thermal expansion is remarkable and the mechanical response significantly influences the neutronics and thermal analyses. Due to the considerable difference between heat pipe cooled reactors and traditional water reactors in the structure and design concept, the coupling solutions for light water reactors cannot be directly applied to heat pipe cooled reactor analyses. Therefore, new coupling framework and program need to consider the coupling effects among neutronics, heat transfer as well as mechanics. Based on the Monte Carlo program RMC and commercial finite element program ANSYS Mechanical APDL, this work introduces the three coupling fields of neutronics (N), thermal (T), and mechanics (M) for heat pipe cooled reactors. The neutronic and thermal-mechanical (N/T-M) coupling strategy is developed theoretically, focusing on the formulation of the nonlinear problem, iteration schemes, and relaxation methods. Besides, the finite element method and the Monte Carlo program use different meshes and geometry construction methods. The spatial mapping and geometry reconstruction are also essential for the N/T-M coupling, which is discussed and established in detail. Furthermore, the N/T-M coupling methods are applied to the preliminary self-designed 10 kWe space heat pipe cooled reactor. Coupling shows that the thermal-mechanical feedback in the solid-state reactor has negative reactivity feedback (-2007 pcm) while it has a deterioration in heat transfer due to the expansion in the gas gap. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Heat pipes

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Connectors based on 2-way SMA actuator for self-reconfigurable chain robots 15

Connectors based on 2-way SMA actuator for self-reconfigurab...

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15th IEEE International Conference on Mechatronics and Automation, ICMA 2018

作者： Ni, Fenglei Li, Yongqiang Zhou, Yunhu Liu, Hong Tang, Yuan State Key Laboratory of Robotics and System Harbin Institute of Technology Harbin Heilongjiang Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781538660720

Self-reconfigurable modular robots have shown great potential in carrying out multiple tasks in different conditions. A connection of both flexibility and reliability is critical for realizing the docking and transformation of the modular robots. We propose a new design of connectors based on SMA which has the advantages of high autonomy, small size and high speed. The couples of connectors consist of active and passive connector. In the active connector, a two-way rotary SMA actuator is used as the actuation mechanism. The double-layer cam and claws are used as an executive mechanism. The passive connector can be securely docked in five different facets for 3D reconfigurations. The multisensory system and hardware architecture provide the foundation for control strategies. The design and the prototype of the connector are described in this paper. Finally, experimental results to evaluate the performance of docking are presented. © 2018 IEEE.

关键词： Modular robots

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Synthesis Optimization of SSZ-13 Zeolite Membranes by Dual Templates for N_(2)/NO_(2) Separation

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Chemical Research in Chinese Universities 2022年第1期38卷 250-256页

作者： LI Jun RONG Huazhen CHEN Cong LI Ziyi ZUO Jiayu WANG Wenjian LIU Xinjian GUAN Yixing YANG Xiong LIU Yingshu ZOU Xiaoqin ZHU Guangshan School of Energy and Environmental Engineering University of Science and Technology BeijingBeijing 100083P.R.China Faculty of Chemistry Northeast Normal UniversityChangchun 130024P.R.China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213P.R.China

High-silica SSZ-13 zeolite membranes are promising in industrial separations of light gases and continuous membranes are highly demanded for better separation performances. Herein, pure-phase, continuous and thin SSZ-13 zeolite membranes were synthesized using dual templates of N,N,N-trimethtyl-1-adamantammonium hydroxide(TMAdaOH) and tetraethylammonium hydroxide(TEAOH). systematical investigations of TMAdaOH/TEAOH ratios and their concentrations show that TMAdaOH acts as the main structure-directing agent in the formation of the SSZ-13 zeolite. TMAdaOH cooperatively plays with TEAOH in promoting the SSZ-13 crystal intergrowth to form a continuous polycrystalline membrane. Additionally, appropriate introduction of TEAOH is able to adjust the membrane thickness to the crystal-comparable size of ca. 2.0 μm. The SSZ-13 membranes are further applied for N2/NO2separation, which is firstly reported on zeolite membranes. The gas permeation results show that the SSZ-13 membrane synthesized by the dual-template approach exhibits selective separation of N2 over NO2 with N2/NO2 separation factor of 7.6 and N2 permeance of 1.66×10-8 mol·m-2·s-1·Pa-1.

关键词： Zeolite membrane Dual template SSZ-13 Gas separation Nitrogen dioxide

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A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

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Nuclear science and Techniques 2022年第2期33卷 53-65页

作者： Fang Zhao Shu-Liang Zou Shou-Long Xu Xuan Wang Jun-Long Wang De-Wen Tang School of Resource Environment and Safety Engineering University of South ChinaHengyang 421001China Hunan Provincial Key Laboratory of Emergency Safety Technology and Equipment for Nuclear Facilities Hengyang 421001China Shanghai Nuclear Engineering Research and Design Institute Co.LTD Shanghai 200000China Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental *** is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear ***,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe *** radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion *** results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity *** the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber *** wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.

关键词： Radionuclide diffusion MELCOR coupled with scSTREAM Severe accident Marine nuclear reactor

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Fault Severity Classification based on Information Flow and Support Vector Machines 8

Fault Severity Classification based on Information Flow and ...

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8th International Conference on Energy system, Electricity, and Power, ESEP 2023

作者： Yang, Xiaohua Liu, Zhikun Fang, Weiyang Huang, Tao Liu, Jie Yang, Jinghua Ren, Changan Department of Compute Science University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (数字)9781510679801

ISBN: (纸本)9781510679795

Fault severity describes the level of impact caused by a fault. In the event of a malfunction in a nuclear power plant, a vast amount of data and alarms are generated, making it challenging for operators to identify and determine the state of the power plant in a short time. If the fault severity can be effectively identified during a malfunction, it would allow the operators to respond quickly and prevent further losses. Compared to other machine learning methods, the Support Vector Machine (SVM) has superior classification capability in scenarios with limited nuclear power samples. However, if the feature representation of the SVM training sample is inadequate, resulting in similar, indistinguishable training data, it will affect the SVM's classification of fault severity. To address this issue, using the Liang-Kleeman Information Flow (IF) can characterize information transmission from a causal perspective. By introducing IF to combine features of the training data, insufficient sample features can be replaced with the IF that describes information from a causal viewpoint, thereby enhancing the SVM's accuracy in fault severity classification. This paper takes the Main Steam Line Break accident in the primary loop of the nuclear reactor as an example and conducts simulation experiments on the PCTRAN simulation platform. The experimental results show that under 5 different fault degrees, the accuracy of fault degree evaluation of SVM model combined with IF is improved. This method can effectively improve the classification accuracy of different fault severity degrees in MSLB steam flow data with similar changes. © 2024 SPIE.

关键词： Steam piping systems

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Condition II Transients Analysis for Pellet Clad Interaction (PCI)

Condition II Transients Analysis for Pellet Clad Interaction...

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The Fifth China-Korea Workshop on Nuclear reactor Thermal Hydraulics

作者： Bi Shumao Shen Caifen Science and Technology on Reactor System Design Technology Laboratory

During an overpower transient, the fuel rod cladding may be subjected to major stresses due to differential expansion of the fuel and the cladding. This may lead to clad failure through the Pellet Clad Interaction (PCI) phenomenon. So the aim of this paper is to present the linear power reached during the most limiting condition II transients(Excessive Load Increase (ELI), Uncontrolled RCCA Withdrawal at Power (URWP) and undetected Rod Drop (RD)) with respect to Qinshan-II long cycle fuel management project PCI analysis.

关键词： ELI PCI Condition II Transients Analysis for Pellet Clad Interaction

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Research of the bundle CHF prediction based on the minimum DNBR point and the BO point methods 16

Research of the bundle CHF prediction based on the minimum D...

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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Liu, Wei Shan, Jianqiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China School of Nuclear Science and Technology Xi'an Jiao Tong University Xi'an710049 China

ISBN: (纸本)9781510811843

There are two ways in the development of CHF correlation in rod bundles: one is based on round tube CHF correlations, considering the effects of spacers and CHF promoters, effects of cold walls, etc. The other one is to develop the correlation directly using the rod bundles CHF experimental data which includes two methods: me minimum DNBR point method and the BO point method. In this study, the local thermal-hydraulic parameters at DNB occurrence point of each set of bundle CHF data were obtained by using the subchannel code ATHAS. Both of the minimum DNBR point and the BO point methods were then applied to develop a new CHF correlation for rod bundles. The so-called "three-step method" was used to determine the shape of me correlation and the non-linear regression analysis method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, a new CHF correlation named the ACC correlation has been developed. The analysis and assessment results indicate that me ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Compared with the minimum DNBR point method, me BO point method is recommended to predict me risk of DNB because it is more reasonable and conservative with better predication rate. Coupled with subchannel code ATHAS, this correlation can simulate the thermalhydraulics performances of rod bundles exactly.

关键词： Forecasting

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Experimental study of thermal and hydraulic characteristics of pressurizertype heating container during blowdown process

Experimental study of thermal and hydraulic characteristics ...

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

作者： Ke, Bingzheng Gao, Puzhen Wang, Bo Chen, Bowen Li, Ru Tian, Ruifeng Cheng, Kun Lu, Chuan Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

Blowdown process is an important process in LOCA. Studying the internal thermal hydraulic characteristics of the pressurizer during LOCA blowdown process can help to accurately analyze the process of the whole LOCA, which is of great significance to reduce the hazard of accidents and improve the safety of nuclear power plants. An experiment system of blowdown process under low pressure conditions is built, and it uses a pressure gauge and a differential pressure level gauge to measure the pressure and liquid level in the pressurizer-type heating container, and uses a thermocouple to measure the temperature in the pressurizer-type heating container at different heights and distances from the centerline. The changes of pressure, liquid level and temperature characteristics during the blowdown process were obtained. The research results show that the higher the heating power, the greater the slope of pressure change in the heating container, and the pressure changes from negative to positive growth;The greater the heating power is, the faster the liquid level decreases, and the liquid level decreases sharply after the blowdown inflection point. The greater the heating power is, the greater the rate of temperature change, from negative growth to positive growth. The temperature change rate after the blowdown inflection point is reduced. As the height increases, the temperature inside the container first increases and then decreases;the further away from the container's centerline, the higher the measured temperature. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Containers

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Modeling of SiC cladding bowing behavior with consideration of spacer grids

Modeling of SiC cladding bowing behavior with consideration ...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Li, Wei Shirvan, Koroush Massachusetts Institute of Technology CambridgeMA02139 United States Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Deep Learning Based Fault Diagnosis for Chemical Process with Statistical Feature Fusion 42

Deep Learning Based Fault Diagnosis for Chemical Process wit...

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42nd Chinese Control Conference, CCC 2023

作者： Zhang, Lei Xia, Hao Cha, Bao School of Control Science and Control Engineering Dalian University of Technology Dalian116024 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610091 China

ISBN: (纸本)9789887581543

The existing fault diagnosis methods based on deep neural networks mostly rely on process data for fault classification and diagnosis, which makes it difficult to distinguish minor faults with similar data features. Therefore, this paper proposes a deep learning-based fault diagnosis model that combines statistical indicators and neural networks. By analyzing and calculating the relevant information entropy to screen the statistical signal features of time domain and frequency domain, we can extract more hidden information from the raw data, and use the convolutional autoencoder to compress the statistical indicators for feature extraction, and use CNN network to extract features from raw data. Finally, the two sets of features are fused and input into the LSTM network to obtain temporal information for fault diagnosis. Tests conducted on the Tennessee Eastman (TE) process show that the proposed model outperforms typical classifiers and improves the diagnostic accuracy of minor faults. © 2023 Technical Committee on Control Theory, Chinese Association of Automation.

关键词： Failure analysis

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