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Research on probabilistic safety assessment (PSA) method of safety class distributed control system (DCS) in nuclear power plant based on fuzzy Markov process

Research on probabilistic safety assessment (PSA) method of ...

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

作者： Tian, Xufeng Hao, Yan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China HuaFu avenue 299# section 3 Chengdu city610213 China

ISBN: (纸本)9780791883778

Safety class DCS(Distributed Control system) in nuclear power plant is characterized by multi-state interaction, complex function and unknown fault mechanism, which limits the use of tradition PSA(Probabilistic Safety Assessment) method such as event tree and fault tree. An improved dynamic probabilistic security simulation method is proposed based on fuzzy Markov process to quantitative evaluation the risk of Safety class DCS in nuclear power plant, which resolved the failure uncertainty problem without in-depth understanding of failure mechanism. Firstly, the fuzzy fault tree reliability evaluation model is established based on fuzzy theory and FTA(Fault Tree Analysis), which describes the probability and degree of failure with fuzzy number. Then, the fuzzy fault tree is transformed into fuzzy Markov chain, this model is described mathematically, and the reliability indexes are simulated and analyzed with Monte Carlo simulation method, the change of system safety performance in long-term operation is obtained. Finally, the scientificity and effectiveness is verified by the example of redundant analog output module of safety class DCS in nuclear plant. Copyright © 2020 ASME

关键词： Nuclear power plants

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

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

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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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Small-Target Infrastructure Detection with Spatial Tracking and Multi-Level Feature Extraction of Railway Track Scenes

Small-Target Infrastructure Detection with Spatial Tracking ...

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Intelligent Communication, Sensing and Electromagnetics (ICSE), International Conference on

作者： Heng Du Huaiming Cang Yiwen Bai Xinyi Wang Beijing Elephant Technology Co. Ltd Beijing China School of Transportation Science and Engineering Beihang University Beijing China The State Key Lab of Intelligent Transportation System Beijing China Tianjin Key Laboratory for Advanced Mechanical System Design and Intelligent Control School of Mechanical Engineering Tianjin University of Technology Tianjin China National Demonstration Center for Experimental Mechanical and Electrical Engineering Education (Tianjin University of Technology) Tianjin China

ISBN: (数字)9798350354935

ISBN: (纸本)9798350354942

In the field of railway transportation, with the continuous improvement of train running speed and the continuous increase of operating mileage, it becomes particularly important to ensure the safety and integrity of small-target infrastructure along the railway train. Although many small-target detection methods have been proposed, the research on small-target infrastructure detection along rail trains is very rare. To solve this problem, this paper proposes a small-target infrastructure detection network with spatial tracking and multi-level feature extraction based on Yolov8n. The experimental results show that ST-YOLO has a 3.6% mAP improvement on the OSDaR23 dataset compared to the baseline Yolov8n, and shows significant advantages over other state-of-the-art methods.

关键词： Rails Transportation Railway safety Switches Feature extraction Rail transportation Real-time systems Sensors Safety Monitoring

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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

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Experimental study on multi-channel jets in plate assembly under blockage condition

Experimental study on multi-channel jets in plate assembly u...

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

作者： Huijian, Huang Peng, Wang Dongyang, Li Tan, Sichao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin Heilongjiang China

ISBN: (纸本)9780791883785

In the nuclear field, the reactor using plate fuel assembly forms multiple jets at exit. So, it's meaningful to study multiple jets to learn the behavior of the coolant when it leaves the core. This paper uses PIV technology to study the flow characteristics of 9 jets at low Reynolds number. Firstly, the velocity field reveals that the pressure outside the jet is greater than the inside of the jet, which causes the jet to deflect and converge inward at z/w=0-8. Then, the velocity field with different Re number is analyzed in the center plane, and find that the flow distribution is similar. Then, a detailed analysis is performed on the jet under specific conditions, and the merge point is discussed in the paper. At the same time, the first-order velocity tensor are also calculated in this paper. In addition, this article also analyzes the jet flow field after the central narrow channel was blocked. In this experiment, a plug was used to block the No. 5 jet by 1/3. And the flow redistribution is discussed in this case. The methods of flow calculation using PIV technology is established and compared with the real number of flowmeters, it's found that the calculation method is rather accurate. On the other hand, the experiment find that the distribution of flow in each channel is not uniform, and the blocking condition make an increase in the flow of edge channels, which leads to the decreasing of flow in the blocked adjacent channel. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Velocity

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Study on natural circulation flow instabilities in rod bundle channel under rolling condition

Study on natural circulation flow instabilities in rod bundl...

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

作者： Cheng, Kun Deng, Jian Cai, Rong Qian, Libo Qi, Peiyao Ke, Bingzheng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Heilongjiang150001 China

ISBN: (纸本)9780791883785

The effects of rolling condition on the flow instability characteristics of natural circulation (NC) in rod bundle channel were experimentally studied. A 3×3 rod bundle channel is used as the testing section. The experimental system pressure range is 0.2 to 0.6 MPa, and the range of inlet subcooling is 10 to 70°C. The ranges of rolling motion amplitude and period are 10 ∼ 20° and 10 ∼ 30s, respectively. Two typical two-phase flow instabilities in rod bundle channel under rolling condition were found in experiments: (a) the trough-type oscillation caused by the vapor generation at the minimum point of flow fluctuation and (b) the compound oscillation formed by the superposition of the trough-type oscillation and DWOI. Experimental results show that the rolling motion can reduce the threshold heating power of trough-type oscillation and cause the occurrence of NC flow instability in advance. But the rolling motion cannot affect the dimensionless boundary of DWOI in rod bundle channel. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Two phase flow

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

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

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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

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

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

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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

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RMC/ANSYS Multi-physics coupling solutions for heat pipe cooled reactors analyses

RMC/ANSYS Multi-physics coupling solutions for heat pipe coo...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Ma, Yugao Liu, Minyun Chen, Erhui Xie, Biheng Chai, Xiaoming Huang, Shanfang Wang, Kan Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781713827245

The heat pipe cooled reactor is a solid-state reactor using heat pipes to passively transfer heat generated from the reactor, which is a potential and near-term space nuclear power system. This paper introduces the coupling scheme between the continuous energy reactor Monte Carlo (RMC) code and the finite element method commercial software ANSYS. Monte Carlo method has the advantages of flexible geometry modeling and continuous-energy nuclear cross sections. ANSYS Parametric design Language (APDL) is used to determine the detailed temperature distributions and geometric deformation. The on-the-fly temperature treatment of cross sections was adopted in RMC code to solve the memory problems and to speed up simulations. This paper proposed a geometric updating strategy and reactivity feedback methods for the geometric deformation of the solid-state core. The neutronic and thermal-mechanical coupling platform is developed to analyze and further to optimize the heat pipe cooled reactor design. The present coupling codes analyze a 2D central cross-section model for MEGAPOWER heat pipe cooled reactor. The thermal-mechanical feedback reveals that the solid-state reactor has a negative reactivity feedback (~1.5 pcm/K) while it has a deterioration in heat transfer due to the expansion. © The Authors, published by EDP sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://***/licenses/by/4.0/).

关键词： Finite element method

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Revealing the Correlation between Microstructure and Fatigue Crack Growth Behaviors in Ti-6al-3nb-2zr-1mo Alloy

SSRN

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SSRN 2022年

作者： Yang, Pengfei Sun, Qi Zheng, Xuechao Mi, Xue Chai, Linjiang Zhu, Minhao Tribology Research Institute School of Mechanical Engineering Southwest Jiaotong University Chengdu China School of Material Science and Engineering Southwest Jiaotong University Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China School of Materials Science and Engineering Chongqing University of Technology Chongqing China

In the current work, the correlation between microstructural features and fatigue crack growth (FCG) behaviors in Ti-6Al-3Nb-2Zr-1Mo alloy is investigated by a combination of electron backscatter diffraction (EBSD), electron channeling contrast (ECC) and mechanical tests. Three kinds of specimens with different microstructures are prepared by water quenching (WQ), air cooling (AC) and furnace cooling (FC) after isothermal holding for 1h in α+β phase region of the as-received material. The da/dN-ΔK curves indicate that the FCG resistance ability of AC specimen is the worst, resulting in the highest FCG rate. The corresponding FCG paths of these specimens are also examined systematically. And the results show that the crack bifurcation and deflection in both WQ and FC specimens are more serious than that in AC specimen. For the specimen with WQ microstructure, boundaries with high angle misorientation angles between different α variants transformed from β phase may act as an obstacle to hinder the dislocation movement and slip transfer, favoring the secondary crack initiation. Similarly, the strain incompatibility generated at the boundaries between different α phases in FC specimen may also promote the occurrence of crack bifurcation. In contrast, in AC specimen, α colonies that contains α variants with the same crystal orientation are more likely to be formed inside β transformation grains leading to few secondary cracks formed. Furthermore, a modified mathematical model is developed to evaluate the co-effect of the crack deflection and bifurcation on the FCG rate. It is suggested that the crack bifurcation is the main factor that can effectively reduce the stress intensity factor of the main crack tip, and in turn slow down the FCG rate. © 2022, The Authors. All rights reserved.

关键词： Titanium alloys

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