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On Dynamic Behavior of Triangular Shaped Cantilevers in MEMS Sensors: Effect of Curvature

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On Dynamic Behavior of Triangular Shaped Cantilevers in MEMS...

2010 10th IEEE International Conference on Solid-state and Integrated Circuit Technology(第十届固态和集成电路技术国际会议 ICSICT-2010)

作者： Guo-Ying Yuan En-Ze Huang Gui-Ming Zhang Zhuang-De Jiang Li-Bo Zhao Yong Li Zhi-Gang Liu State Key Laboratory for Manufacturing Systems Engineering Xi'an Jiaotong University Xi'an 710049 State Key Laboratory for Manufacturing Systems Engineering Xi'an Jiaotong University Xi'an 710049 State Key Laboratory of Tribology Tsinghua University Beijing 100084 China State Key Lab of Multiple Flow in Power Engineering Xi'an Jiaotong University Xi'an 710049 China

ISBN: (纸本)9781424457977

A study on the effect of curvature of triangular shaped cantilevers in MEMS sensors was presented. 3-D models were performed by using the SolidWorks software. And modal analysis and harmonic response analysis were carried out on different types of geometrically configured triangular shaped cantilevers using ANSYS software. The results included the variation of natural frequencies of cantilevers and maximum equivalent von mises strain on the cantilevers with respect to radius of curvature of triangular shaped cantilevers. According to the finite element analysis results, the curvature did affect dynamic behavior of triangular shaped cantilevers in MEMS sensors. Usually, the curvature is induced by the surface stress due to microprocessing or any change of temperature, etc. Therefore, in order to achieve a steady state of strain for the MEMS sensors, it's better to eliminate the surface stress and make the cantilever transform from a curved cantilever to a straight cantilever.

关键词： CANTILEVER Curvature Surface stress Triangular Microelectromechanical systems Sensors Microtechnology Response analysis

Development and Verification of Safety Analysis Code to Model and Simulate Steam power Conversion System in Sodium-Cooled Fast Reactor

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SSRN 2023年

作者： Liu, Shuo Jin, Zhao Wang, Chenglong Zhang, Dalin Tian, Wenxi Qiu, Suizheng Su, Guanghui Yang, Jun School of Nuclear Science and Technology State Key Lab. of Multiphase Flow in Power Engineering Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China China Institute of Atomic Energy Beijing102413 China

Sodium-cooled fast reactor (SFR) is one of the most promising generation IV reactor type. SFR system is composed of three heat transfer circuits. SFR will generate electricity, which requires accuracy prediction of steam power conversion system. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR has been developed. A modular modeling method was applied. Each equipment is packaged as a separate module. The system model can be obtained by connecting modules to achieve flexible modeling. Compared with RELAP5, the code uses a general algorithm to solve ODE, which enhances the universality of code. The model is verified with design results of Daya Bay Nuclear power Plant and BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. It is proven that the code developed can be used in the analysis and design the steam power conversion system. The steady-state analysis of SFR steam power conversion circuit is carried out. The design could provide a reference for the future development of steam power conversion system in SFR. © 2023, The Authors. All rights reserved.

关键词： Sodium-cooled fast reactors

flow Fluctuations and flow Friction Characteristics of Lead-Bismuth Eutectic in a Vertical Tube Under Rolling Conditions

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SSRN 2022年

作者： Liu, Zhipeng Zhao, Yafeng Tian, Wenxi Wang, Chenglong Qiu, Suizheng Su, Guanghui H. School of Nuclear Science and Technology Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China State Key Laboratory of Multiphase Flow in Power Engineering China

The effects of rolling motion on the flow characteristics of lead-bismuth eutectic in a vertical circular tube are investigated experimentally. The results show that the pressure drop and flow rate of the forced LBE circulation fluctuate under rolling conditions. The effect of rolling motion on the flow characteristics decreases with the increase of the corresponding static Reynolds number. The rolling motion may increase the time-averaged frictional resistance coefficient of the LBE when flow rate is low. The critical value of the impact Reynolds number is about 125,000. As a result, the time-averaged flow rate decreases. The analysis shows that it is mainly the transverse additional force caused by the rolling motion increases the frictional resistance of the LBE. An empirical correlation is established to predict the frictional resistance coefficient under rolling conditions. This study may be useful in the thermal-hydraulic analysis of LBE as a working fluid under moving conditions. © 2022, The Authors. All rights reserved.

关键词： Reynolds number

Steady-state Multi-Physics Coupling Analysis of Heat Pipe Cooled Reactor Core

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SSRN 2022年

作者： Zhang, Zhipeng Wang, Chenglong Guo, Kailun Tian, Wenxi Qiu, Suizheng Su, Guanghui Department of Nuclear Science and Technology Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China State Key Laboratory of Multiphase Flow in Power Engineering China

Due to the advantages of mobility, safety and universality, the heat pipe reactors have great competitiveness in many applications such as strategic defense locations, remote communities, and emergency locations. In addition, most of the heat pipe reactors are designed as solid-state reactors, which avoid the two-phase flow problem caused by the coolant transformation and make it possible to analyze the reactor with multiple physical field coupling. In this work, a typical heat pipe reactor has been carried out a 3-D neutronic-thermal-mechanical analysis, and steady state analysis is performed by OpenMC coupled with COMSOL. Coupling calculation results show that under normal operation, the maximum temperature and thermal expansion of core is 1338.94 K and 2.53 mm. In the steady-state coupling process, the influence of temperature parameters on physical calculation is mainly reflected in k eff and its power level, the maximum temperature drops from 1347.26 K to 1338.94 K, but it has little influence on power distribution. By coupling calculations, the maximum stress of the reactor is reduced by 5.78 MPa. The research results will support the subsequent development of multi-physics coupled transient analysis code for heat pipe reactors. © 2022, The Authors. All rights reserved.

关键词： Heat pipes

Dynamic Modeling of Steam power Conversion System in Sodium-Cooled Fast Reactor

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SSRN 2023年

作者： Liu, Shuo Jin, Zhao Wang, Chenglong Zhang, Dalin Tian, Wenxi Qiu, Suizheng Su, G.H. Yang, Jun School of Nuclear Science and Technology State Key Lab. of Multiphase Flow in Power Engineering Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China China Institute of Atomic Energy Beijing102413 China

Sodium-cooled fast reactor (SFR) is the most promising generation IV reactor type. SFR systems are composed of three heat transfer circuits. Currently, research on steam power conversion system is scarce because of complex structure. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR has been developed. Models of main components have been established, including once-through steam generator (OTSG), turbine, deaerator, feedwater heater, condenser, pump and moisture separator reheater (MSR). The model is verified with design results of Daya Bay Nuclear power Plant and BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. It is proven that the code developed can analysis and design the steam power conversion system. The steady-state analysis of SFR steam power conversion circuit is carried out. The design could provide a reference for the future development of steam power conversion system in SFR. © 2023, The Authors. All rights reserved.

关键词： Sodium-cooled fast reactors

Highly accurate meshless method for convection-diffusion problems

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Kung Cheng Je Wu Li Hsueh Pao/Journal of engineering Thermophysics 2008年第9期29卷 1561-1563页

作者： Wu, Xue-Hong Li, Zeng-Yao Ma, Liang-Dong Tao, Wen-Quan State Key Lab. of Multiphase Flow in Power Engineering Xi'an Jiaotong University Xi'an 710049 China

Meshless local Petrov-Galerkin method (MLPG) as a new computational method has been developed in recent years, which disposes a group of scatter points in the computational domain and construct the interpolation function through these nodes. The Smith-Hutton problem was selected as the demonstrated test problem and solved by the MLPG. The results of the MLPG method were compared with the results of finite volume method with high order scheme. The results indicate that the MLPG method is a highly accurate numerical method and can deal effectively with high Pe number fluid flow problems.

关键词： Finite volume method

Experimental Investigation on flow and Heat Transfer Characteristics of Helium in Rectangular Narrow Slit Channel

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SSRN 2024年

作者： Li, Changzhong Wang, Chenglong Sun, Yanyu Chen, Ronghua Tian, Wenxi Su, Guanghui Qiu, Suizheng State Key Laboratory of Multiphase Flow in Power Engineering China Department of Nuclear Science and Technology Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China China Nuclear Power Engineering Co. Ltd. Beijing100840 China

Helium is utilized as coolant in high temperature gas cooled reactors. Rectangular narrow slit channels are commonly present in the core. The flow and heat transfer properties of helium significantly affect the core temperature and flow distribution. This paper experimentally investigates the flow and heat transfer properties of high temperature helium in rectangular narrow slit channel. The Reynolds numbers ranged from 468 to 5327, the temperature ratio of wall to bulk from 0.91 to 1.22, the helium outlet temperature up to 944 K, and the maximum heat flux density up to 0.101 MW/m². In the experiment, the total and local convective heat transfer coefficients, along with the friction factors, are measured. Correlation equations for the Nusselt number and friction factor with respect to Reynolds number are derived. The primary findings are as below: the friction factors of helium in rectangular narrow slit channels are significantly larger than those calculated by current empirical correlations. In the turbulent zone, the measured local Nusselt numbers are in excellent accordance with the Gnielinski correlation;however, a significant discrepancy exists in the laminar zone. New flow heat transfer correlation equations are proposed on the bases of experimental data. Upon comparison, the total Nusselt numbers fall in the error of 10%, and almost all local Nusselt numbers and the friction factors fall in the error of 20%. © 2024, The Authors. All rights reserved.

关键词： Helium

Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly

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Experimental Thermal and Fluid Science 2018年 96卷 180-191页

作者： Jiangtao Yu Dalin Zhang Leitai Shi Zhiwei Wang Wenxi Tian G.H. Su S.Z. Qiu State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University 28 Xianning West Road Xi’an 710049 China Shaanxi Key Lab. of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University 28 Xianning West Road Xi’an 710049 China

To investigate Counter-Current flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors.

关键词： Steam-water counter-current flow Partial CCFL correlation Pressurizer surge line assembly Advanced passive nuclear reactors ADS Automatic Depressurization System ADS1-3 First to Third Stages of ADS ADS4 Fourth Stage of ADS CCF Counter-Current flow CCFL Counter-Current flow Limitation CCFL-1 The 1st grade of CCFL region CCFL-2 The 2nd grade of CCFL region CCFL-3 The 3rd grade of CCFL region CCFL-L Local CCFL at the T-junction of HL pipe and PZR SL pipe CCFL-S Local CCFL along the PZR SL pipe CCFL-U Local CCFL at the location between the PZR simulator bottom head and vertical part of PZR SL pipe CCFL-ZP Zero liquid penetration region or stage CMT Core Makeup Tank COCF Co-Current flow HL Hot Leg IRWST In-containment Refueling Water Storage Tank Ku Kutateladze PRHR Passive Containment Cooling System PWRs Pressurized Water Reactors PZR SL Pressurizer Surge Line RC Reactor Core RCS Reactor Coolant System RPV Reactor Pressure Vessel SBLOCAs Small-Break Loss-of-Coolant Accidents ZP zero liquid penetration

Nuclear Thermal Coupling Optimization of Baffled Ntp Reactor

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SSRN 2022年

作者： Duan, Zimian Zhang, Jing Han, Zichao Wang, Mingjun Wu, Yingwei He, Yanan Tian, Wenxi Qiu, Siuzheng Su, Guanghui Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Xi’an710049 China Department of Nuclear Science and Technology State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University Xi’an710049 China

Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactor has received a lot of research due to its simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo neutron dynamics program and commercial CFD program Fluent. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than $3.44 and a minimum dry shutdown reactivity lower than -$5.95, where there was a high control accuracy of 6.7 pcm/mm. The reactor realized safety of the core temperature lower than 3000K under the conditions of total exit temperature of 2650K and thrust of 25000lb. © 2022, The Authors. All rights reserved.

关键词： Cermets