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Modal Analysis of Lead-Bismuth Eutectic Flow in a Single Wire-Wrap Rod Channel

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SSRN 2022年

作者： Zhao, Xielin Cheng, Qian Yu, Xiaofei Huang, Qian Zhang, Ke Feng, Zhipeng Zhou, Jinxiong State Key Laboratory for Strength and Vibration of Mechanical Structures School of Aerospace Xi'an Jiaotong University Xi'An710049 China Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu610200 China

The thermal-hydraulics as well as flow-induced vibration of wire-wrapped rod bundles calls for accurate and efficient liquid metal flow simulation and prediction, yet it remains a challenge due to the complex geometries and high Reynolds number flow in wire-wrapped rod channel. Previous efforts towards this goal exclusively adopts full-order modeling (FOM), which is prohibitively computation-intensive. This work reports the first reference on modal analysis of the turbulent lead-bismuth eutectic (LBE) flow in a single wire-wrapped rod channel by a data-driven dynamic mode decomposition (DMD) method. The spatio-temporal data of fluid velocities and pressure were modelled and collected by using the Reynolds-averaged NavierStokes equations (RANS) approach with stress-omega RSM turbulence model. Model analysis was performed over the snapshot matrices and DMD modes and corresponding eigenvalues were calculated. The DMD modes and eigenvalues were then used to construct a reducedorder modeling (ROM) system for fast fluid field reconstruction and short-term forecasting. The correctness and feasibility of the method was demonstrated and compared with groundtruth CFD simulation results. Our method offers a fast and reliable approach for ROM of turbulent flow in wire-wrapped rod fuel assemblies. We publicize all the data and codes via https://***/XJTU-Zhou-group/Wire-wrapped-fuel-pin-CFD-DMD. © 2022, The Authors. All rights reserved.

关键词： Dynamic mode decomposition

Thermionic Conversion Performance Analysis of the Single-Cell Thermionic Fuel Element Based on Froba-Thermion Code

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SSRN 2022年

作者： Lu, Kailin Yao, Hao He, Yanan Wu, Yingwei Zhang, Jing Liao, Haoyu Su, Guanghui H. Qiu, Suizheng 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 Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

The thermionic conversion performance of the single-cell thermionic fuel element (TFE) is influenced by fuel mass transfer and other parameters. With the consideration of current flow and heat transfer, the thermionic conversion model and the modified heat transfer model were implemented in FROBA-THERMION. Afterward, both implemented models were verified by comparing with experiments or KATET calculation results. Further, a steady-state performance simulation of the single-cell TFE was carried out. The results indicate that the thermionic conversion efficiency will be decreased due to the emitter temperature flattening caused by fuel mass transfer. In addition, the influence of several key parameters on thermionic conversion efficiency and measures to improve the thermionic conversion efficiency were analyzed. © 2022, The Authors. All rights reserved.

关键词： Mass transfer

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

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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.

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Effect of yttrium and calcium additions on electrochemical behaviors and discharge performance of AZ80 anodes for Mg-air battery

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Transactions of Nonferrous Metals Society of China 2022年第8期32卷 2510-2526页

作者： Yu-wen-xi ZHANG Lu HAN Lin-bao REN Ling-ling FAN Yang-yang GUO Ming-yang ZHOU Gao-feng QUAN Key Laboratory of Advanced Technologies of Materials(Ministry of Education) School of Materials Science and EngineeringSouthwest Jiaotong UniversityChengdu 610031China Center for Advancing Materials Performance from the Nanoscale State Key Laboratory for Mechanical Behavior of MaterialsXi’an Jiaotong UniversityXi’an 710049China Xi’an Rare Metal Materials Institute Co. Ltd.Xi’an 710049China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

The effects of yttrium(Y)and yttrium+calcium(Y+Ca)additions on the electrochemical properties and discharge performance of the as-extruded Mg−8Al−0.5Zn−0.2Mn(AZ80)anodes for Mg−air batteries were *** results show that the addition of 0.2 wt.%Y increased the corrosion resistance and discharge activity of AZ80 *** was attributed to the fine and sphericalβ-Mg_17)Al_(12) phases dispersing evenly in AZ80+0.2Y alloy,which suppressed the localized corrosion and severe“chunk effect”,and facilitated the rapid activation ofα-*** addition of 0.2 wt.%Y and 0.15 wt.%Ca generated grain refinement and a reduction of theβ-Mg_17)Al_(12) phase,resulting in a further enhancement in discharge ***,the incorporation of Ca in Mg_17)Al_(12) and Al_(2)Y compounds compromised the corrosion resistance and anodic efficiency of AZ80+0.2Y+0.15Ca ***,AZ80+0.2Y anode exhibited excellent overall discharge performance,with the peak discharge capacity and anodic efficiency of 1525 mA·h·g^(−1) and 67%at 80 mA/cm^(2),13%and 14%higher than those of AZ80 anode,respectively.

关键词： Mg-air battery Mg-Al-Zn anode discharge performance electrochemical behavior

design of an Online Operational Task Reliability Analysis system in Nuclear Power Plant

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Design of an Online Operational Task Reliability Analysis Sy...

system Reliability and Safety Engineering (SRSE), International Conference on

作者： Wei Li Ming Yang Zhengxi He Zhao Gao Yan Xiong Zhongyue Zhang Dehang Jiang Jipu Wang Fundamental Science on Nuclear Safety and Simulation Technology Laboratory College of Nuclear Science and Technology Harbin Engineering University Harbin China Nuclear Reactor System Design Technology Key Laboratory China Nuclear Power Research and Design Institute Chendu China Institute for Advanced Study in Nuclear Energy & Safety College of Physics and Optoelectronic Engineering Shenzhen University Shenzhen China

Human factors play an important role in human- computer interaction systems such as nuclear power plants (NPPs). Especially for the human-computer interaction process in unknown scenarios, incorrect personnel actions will affect the operation of equipment and systems, and finally the achievement of safety goals of NPPs. This paper presents a design of an online operational task reliability analysis (OOTRA) system which provides early warnings to operations in a timely manner by monitoring the states of plant systems and equipment, and operator actions. For any wrong action of operators, the OOTRA system can online evaluate the impact of the wrong action on the operational task and the goals of NPPs.

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

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

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

Study on Imc-Pid Control of Once Through Steam Generator for Small Lead Bismuth Fast reactor

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SSRN 2022年

作者： Sun, Aodi Zhang, Ru Xiao, Kai He, Zhengxi Sun, Peiwei Wei, Xinyu Key Laboratory of Advanced Nuclear Energy and Technology Shaanxi Engineering Research Center of Advanced Nuclear Energy Xi’an Jiaotong University Xi’an710049 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Due to the particularity of nuclear reactor, it is difficult to adjust the controller parameters by experiment. The simulation calculation inevitably simplifies the actual model. The controller parameters tuned by this model is likely to cause poor performance due to model mismatch. Therefore, the research on the adaptability of the controller to model mismatch has guiding significance for engineering application. In this paper, a once through steam generator(OTSG) for small lead bismuth fast reactor(SLBFR) is established and verified, and the presented model is simplified by three different methods. Internal model control(IMC) and IMC-PID controllers are designed based on the simplified models. With the increase of λ , IMC-PID controller becomes insensitive to model mismatch caused by model linearization, non-minimum phase characteristics, noise and time delay, and has high robustness to model mismatch. However, the robustness to model sacrifices the sensitivity of the system. In the study of rapid power reduction, it is found that large λ will lead to conservative control output, and then make the control performance worse. Through the research of this paper, Method 1 is recommended, and λ is selected between 5 and 10 to design feedwater IMC-PID controller for OTSG of SLBFR. © 2022, The Authors. All rights reserved.

关键词： Model predictive control

Development of the Expanding Mandrel Test Facility for Investigation of Zircaloy Cladding Performance Under Loading Strain Rate Similar to Power Ramp

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SSRN 2022年

作者： He, Yanan Wu, Yingwei Qiu, Bowen Lei, Zhao He, Liang Jiao, Yongjun Chen, Ping Tian, Wenxi Su, Guanghui Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology Shaanxi Engineering Research Center of Advanced Nuclear Energy Xi’an Jiaotong University Xi'An710049 China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

The in-pile pellet-cladding interaction (PCI) is the one of challenging problems for conventional UO2-Zicaloy fuel systems, which can threaten the integrity of the fuel rod and lead to the release of fission product. Currently, several experiments and numerical studies have been devoted to PCI analysis. However, attributing the lacking of experimental data with high fidelity, the development, and validation of the PCI analysis model is not sufficient. In this study, an expanding mandrel test facility has been developed to investigate PCI behaviors, where the digital image correlation (DIC) measuring system was introduced to online monitor the displacements/strain of cladding outer surface with improved fidelity. In addition, the oxygen content measurement system was employed to identify the failure of cladding. Subsequently, the expanding mandrel test facility was leveraged to investigate the PCI performance of three types of cladding (e.g. N36 cladding, recrystallized Zircaloy4 cladding, and M5 cladding) under both with/without iodine corrosion and cladding real-time strain, failure time and failure strain were recorded. Then, performance comparisons among these claddings were completed and discussed. The results show that the established facility can be applied to expanding mandrel tests under varied theoretical loading strain rates (10~60%/h) and the working environment temperature can reach 350 . Meanwhile, there are three distinct stages for the deformed Zicaloy cladding during the expanding mandrel test. The comparison results suggest that recrystallized Zircaloy4 cladding shows the largest cladding failure strain with/without iodine corrosion under 20%/h theoretical loading strain, while M5 cladding and N36 cladding present considerable performance without iodine corrosion effect. In addition, N36 and recrystallized Zircaloy4 cladding seems more iodine corrosion resistant than that for M5 cladding, which can be beneficial for its in-pile performance. Besides,

关键词： Corrosion resistance

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

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

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