检索结果-内蒙古大学图书馆

Fracture mechanics analysis of cracked structures using weight function and neural network method

学校读者我要写书评

暂无评论

IOP Conference Series: Materials science and Engineering 2018年第1期372卷

作者： J G Chen F G Zang Y Yang K K Shi X L Fu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China China

Stress intensity factors(SIFs) due to thermal-mechanical load has been established by using weight function method. Two reference stress states sere used to determine the coefficients in the weight function. Results were evaluated by using data from literature and show a good agreement between them. So, the SIFs can be determined quickly using the weight function obtained when cracks subjected to arbitrary loads, and presented method can be used for probabilistic fracture mechanics analysis. A probabilistic methodology considering Monte-Carlo with neural network (MCNN) has been developed. The results indicate that an accurate probabilistic characteristic of the KI can be obtained by using the developed method. The probability of failure increases with the increasing of loads, and the relationship between is nonlinear.

关键词：

Numerical Simulation of Flow-Thermal Coupling Field Inside Upper Plenum and Hot Legs of Pressurized Water reactor (Pwr)

学校读者我要写书评

暂无评论

SSRN 2022年

作者： Sun, Ziyun Zhou, Xinzhi He, Zhengxi Xu, Tao Zhu, Jialiang Dong, Chenlong College of Electronics And Information Engineering Sichuan University No. 24 South Section 1 1st Ring Road Sichuan Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No. 328 Section 1 Changshun Avenue Sichuan Chengdu China

The temperature and flow rate of coolant in the hot legs of pressurized water reactor (PWR) systems directly reflect the nuclear power and the heat transfer state of the reactor core, and are key parameters for reactor power control and safety protection. Most of existing studies have been conducted to analyze the steady-state thermal hydraulic features in the upper plenum and hot legs from a spatial scale, but less from a temporal scale to describe and analyze the transient features. In order to meet the requirements of key parameters measurement and safe operation, this paper established a reasonably simplified 3D geometrical model of the upper plenum and hot legs of the HPR1000 reactor based on a multi-physics field simulation platform and investigated the steady-state and transient features of the flow-thermal coupling field of coolant under numerical simulation, focusing on the coupling analysis of the flow-thermal coupling field on the spatial and temporal scales. The analysis of the steady-state features shows that the coolant flowing into the upper plenum from the upper core plate is characterized by a high middle temperature and a low peripheral temperature. As a result, there is an uneven coolant temperature distribution at the inlet of the hot legs that the temperature differences between the cold lower part and hot upper part are relatively large. However, with the axial flow of coolant, the temperature distribution in hot legs tends to be more uniform and the temperature differences decrease. The simulation results show that the flow of the low temperature coolant in hot legs dominates the change of coolant temperature distribution. The analysis of the transient features shows that a small fluctuation of flow rate at the core outlet causes changes in the turbulent kinetic energy of coolant in hot legs, but the effect on coolant temperature distribution is relatively weak. These such variations on the temporal scale can change the coupling relations of c

关键词： Multiphysics

Development, Validation and Application of Sub-Channel Thermal-Hydraulic Analysis Code for Liquid Metal Cooled reactor

学校读者我要写书评

暂无评论

SSRN 2024年

作者： Wang, Jinshun Chen, Ronghua Liang, Yu Gui, Minyang Zhu, Xinyang Tian, Wenxi Qiu, Suizheng Su, Guanghui School of Nuclear Science and Technology Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University Xi’an710049 China Nuclear Power Institute of China 328 Changshun Avenue Chengdu610213 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China College of Nuclear Science and Technology Harbin Engineering University Harbin150001 China

The liquid metal reactor, specifically the lead-based fast reactor and the sodium-cooled fast reactor, are generally recognized as the most significant representatives of the Gen IV reactors due to their sustainability, economy, security and reliability. In order to conduct thermal-hydraulic safety analysis for liquid metal cooled fast reactors, adaptive improvements are made to the model and solution algorithms based on the sub-channel analysis method, and a Sub-channel Analysis Code Of Safety for Liquid Metal cooled reactor (SACOS-LMR) is developed. The validation of SACOS-LMR is conducted using classic experimental data of liquid metal. The results show that the dimensionless temperature deviations of ORNL 19-pin test and Toshiba 37-pin test are within 10%, and the relative temperature deviations of KIT-KALLA 19-pin test are within 25%. These findings confirm the accuracy and reliability of SACOS-LMR in predicting the temperature distribution of liquid metal reactors. Ultimately, calculations conducted on the ALFRED lead-cooled fast reactor demonstrated the practical utility of the code and have certain guiding significance for its subsequent design and development. © 2024, The Authors. All rights reserved.

关键词： Liquid metals

The study of radiation effect on Raman Distributed optical fiber Temperature Sensor

学校读者我要写书评

暂无评论

Journal of Physics: Conference Series 2021年第1期1780卷

作者： Meiqiong Xiang Yanyang Liu Xianguo Qing Zhengxi He Jialiang Zhu Peng He XinLv Science and Technology on Reactor System Design Technology LaboratoryNuclear Power Institute of China Chengdu610213China

The rapid development of big data and artificial intelligence technology is inseparable from the acquisition of massive basic data, and the data is acquired by sensors. The Raman Distributed optical fiber Temperature Sensor (RDTS) is a kind of sensor with many advantages, such as: anti-electromagnetic interference, intrinsic safety, small size, water and high temperature resistance. It has been applied in many fields for fast and continuous temperaturemeasurement. However, due to the rapid attenuation of transmitted light and the temperature error in radiated environment, it has not been applied innuclear field. Therefore, in order to improve the temperature accuracy of RDTS in radiated environment and provide basic data for advanced control. This paper studies the radiation effect on RDTS. Analyzesthe radiation damage mechanism and radiation damage influencing factors of RDTS. Proposes measures to improve the anti-radiation performance and provides four useful calibration methodsfortemperature error. Among them, two calibration methods have been proved right by experiments.

关键词：

design of Shocks & Vibrations Monitoring system for Long Distance Transportation in Confined Spaces

学校读者我要写书评

暂无评论

Design of Shocks & Vibrations Monitoring System for Long Dis...

Chinese Automation Congress (CAC)

作者： Shiyong Huang Yuru Geng Yong Han Mingfei Gu Chengdu China Nuclear Haichuan Nuclear Technology Co. Ltd Nuclear Power Institute of China Chengdu China School of Network and Communication Engineering Chengdu Technological University Chengdu China Chengdu Vocational & Technical College of Industry Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

In real life, many valuable products and highvalue important products are transported in long distances. Shocks and vibrations are caused during transportation, loading and unloading for various reasons. Any shock and vibration exceeding the regulations may cause damage to the transported products. So, it is necessary to monitor and analyze these data of shocks and vibrations scientifically, while considering that some transportation conditions have special requirements for installation space and transportation time. Therefore, an ultra-thin and long-lasting monitoring system has been developed based on MEMS accelerometer technology to measure shocks and vibrations. After series of independent tests and calibration tests conducted by a third party on the product, the results show that it can record and store the threeaxis acceleration, temperature, humidity, pressure, time and other parameters of the transportation process in real time, and complete the preliminary assessment of the transportation status of high-value products. Then engineers can conduct professional analysis of the data to obtain scientific transportation monitoring reports, and provide scientific basis for taking safety prevention and treatment measures for subsequent transportation.

关键词：

Thermal-Mechanical Influence Aspects and Evaluation of Helical Cruciform Single Rod in Fluoride-Salt-Cooled High-Temperature Advanced reactor

学校读者我要写书评

暂无评论

SSRN 2024年

作者： Chen, Yiwen Zhang, Dalin Jiang, Dianqiang Li, Wei Wu, Xiaoli Tian, Wenxi Qiu, Suizheng Su, Guanghui Lu, Qi 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’an710049 China State Key Laboratory for Strength and Vibration of Mechanical Structures Xi’an Jiaotong University Xi’an710049 China Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China China

Helical cruciform fuels are novel in nuclear reactors, potential to increase reactor’s power density. However, the geometry is complicated so influence of it on the fuel performance is not identified yet. In this paper, impact of power density, cross-section parameters, and twist pitch on thermal and mechanical characteristics of fuel used in Fluoride-Salt-cooled high-Temperature Advanced reactor is discussed based on fluid-thermal-mechanical coupling. For thermal features, twist pitch is key while others have few effects. Increase in twist pitch leads to temperature uniformity and rise owing to weaker mixing effects. Fuel center temperature at middle plane of 300mm-pitch rod is 829.58℃, 54.66℃ higher than that of 100mm-pitch rod. Difference of maximum and minimum temperature at 45° cross section of 500mm-pitch rod is 153.62℃, while the figure is only 74.00℃ of 300mm-pitch rod. Besides, axial temperature of cladding outer surface increases wavelike due to complex geometry. For mechanical features, not geometry sizes but temperature affects stress distribution. Maximum Von-Mises stress appears at the elbow, where maximum temperature exists, 116.8MPa under normal conditions, lower than tensile strength of the material. After identifying the influence of these factors, five dimensionless parameters are proposed to evaluate and rate fuel performance based on Technique for Order Preference by Similarity to an Ideal Solution. As a result, an optimization comes up, scoring 0.310, twice more than the original design, owing to the thermal uniformity and mechanical safety. This study provides a reference for identifying and evaluating the performance of helical structure. © 2024, The Authors. All rights reserved.

关键词： Geometry

The Instability Marsh Phenomenon and Marginal Stability Boundary Distortion of Density Wave Oscillation in Parallel Channels

学校读者我要写书评

暂无评论

SSRN 2024年

作者： Sun, Ducheng Ma, Zaiyong Lu, Q. Liu, C.W. Pan, L.M. School of Energy and Power Engineering Shandong University No. 17923 Jingshi Road Jinan250061 China Key Laboratory of Low-grade Energy Utilization Technologies and Systems Ministry of Education Chongqing400044 China Department of Nuclear Engineering and Technology Chongqing University Chongqing400044 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

Density wave oscillation (DWO) in parallel channels is an important problem that has been widely studied. The marginal stability boundary (MSB) is often found to be "L " shaped on the N pch -N sub plane, but some very different shapes were also reported. In this paper, the instability marsh region which could occur in the traditional instable region but having both instability and stability sections was presented and discussed, which is found to be an explanation of the distortion of MSB. The analysis of the instability marsh region showed that the evolution of instability marsh should be due to the stability change when operation condition changes, which can be estimated with the analysis of the two-phase and single phase pressure drop variation, as well as Ma’s stability criterion for different heat flux profiles. Variation of instability marsh with different heat flux profile, inlet and outlet resistance coefficient, mass flow and pressure were obtained and analyzed, and the effects of these parameters on instability marsh variation were found to be well interpreted with stability change. The calculation and analysis of instability marsh showed a clear evolution figure of the distortion of MSB under different conditions, and related research could be helpful in better understanding the flow instability in parallel channels. © 2024, The Authors. All rights reserved.

关键词： Stability criteria

Evaluation and Analysis of RT Suspected Defects on reactor Pressure Boundary Weld

学校读者我要写书评

暂无评论

IOP Conference Series: Earth and Environmental science 2021年第3期632卷

作者： Bo Yang Qiang Deng Cong Chen FengShou Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu China

The control rod driving mechanism (CRDM) is one of the important equipments in the nuclear reactor, and its sealing shell is part of the pressure boundary of the nuclear reactor. The sealing shell is a pressure vessel made of two different material (austenitic stainless steel and nickel base alloy) cylinders welded by dissimilar metal welds. Product production process, in this dissimilar metal weld ray detection found suspected defects show. In this paper, the mechanism and process of the suspected defects are expounded through experimental verification, and it is proved that the suspected defects do not affect the quality of the weld.

关键词：

Limit load estimation for finite plates with semi-elliptical defects under tensile loading

学校读者我要写书评

暂无评论

Journal of Physics: Conference Series 2021年第1期1777卷

作者： K K Shi X L Fu H Xie X J Shao J Tian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

The ASME code allows the designer to circumvent many of the problems arising in design by elastic analysis by performing plastic or limit analysis of the component. Unlike elastic analysis, these analyses take account of the redistribution of stress upon yield. In limit analysis, the calculation of the limit load for finite plate with different semi-elliptical defects is an important stage for mechanical safety evaluation. The elastic compensation technique proposed by Mackenzie can be used to provide the limit load of a structure with a finite element method and aims at producing a suitable stress filed for the lower bound theory. In the present study, it is assumed that the finite plate exists three kinds of semi-elliptical cracks with low aspect ratio a/c (a/c=0.1, 0.5 and 1.0), respectively. The improved elastic compensation technique realized by finite element method is applied in the limit load estimation. At last, the results show that the improved elastic compensation technique can be used to obtain the limit value for the finite plate with different semi-elliptical defects under tensile loading. Moreover, the improved elastic compensation technique can be easily implemented in any finite element method and enables the limit load for any complex structure to be obtained swiftly.

关键词：