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检索条件"机构=Science and Technology on Reactor System Design Technology Laboratory"

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reactor temperature control system based on multi-model predictive control 27

Reactor temperature control system based on multi-model pred...

引用

27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Yun, Yu Jian, Sun Liu, Hongchun Pan, Zhu Zhou, Jixiang Qi, Ye Science and Technology on Reactor Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

reactor coolant temperature control plays a central role among all reactor control systems because the effect of temperature control is directly related to the load tracking capability and also to the reactor protection margins, such as the over-temperature and over-power restraints. Special for the Mode-A reactor, the capacity of reactor is depending on the accurate and quality of temperature control. For example, in most of the time, all the Mode-A reactors in China are operating at basic load, so the temperature control is always worked in an simple way. But the temperature control system must also consider the quicker response and capacity to the wide range change of load, this paper based on point-kinetic model and heat transfer model in reactor, designs a suitable temperature model for control system. Considering that the traditional PID control method is difficult to solve the robustness and stability of high-order systems for large range operation, this paper uses Multi-model Predictive Control to solve this problem. According to the mechanism model, and establish nine single models, this paper online calculation model -matching degree to get recursive bayesian probability weighting, then get global prediction model. According to the global model, design Multi-model predictive control strategy. The simulation results indicate that under large conditions change, also robust and stable control performance is obtained. Copyright © 2018-2019 by JSME

关键词： Model predictive control

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Application of STPA to the digital reactor protection system in NPP for system safety analysis 23

Application of STPA to the digital reactor protection system...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Liu, Zhaohui Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

In NPP, the digital control system which integrated software and hardware are increasingly used to improve dependability and introduce new functionality. Traditional safety analysis can get a good result when handling accidents caused by component failures, but software does not fail in this way. STPA is a new hazard analysis technique based on systems theory rather than reliability theory. It considers the system as a whole (include the hardware and software) to analyze failure and causality of systems and treats safety as a control problem rather than a failure problem. Being a safety-critical system, RPS in NPP needs to be considered carefully in system safety. So, we adopt this new approach to analyze the design process. From the analysis results, we found that causal factors leading to safety accidents identified by STPA included all the hazards identified by the fault tree analysis. Furthermore, there are some causal factors that were identified by STPA only. We utilize these results of the analysis on causation factor to refine the safety requirements and reduce the occurrences of the hazardous scenarios. Copyright © 2015 by JSME.

关键词： Fault tree analysis

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Meso-macro scale model for dynamic fracture of functional graded material 14

Meso-macro scale model for dynamic fracture of functional gr...

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14th International Conference on Fracture, ICF 2017

作者： Bai, Xiaoming Shi, Kaikai Guo, Licheng Zheng, Liangang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Department of Astronautic Science and Mechanics Harbin Institute of Technology Harbin150001 China

A meso-macro scale model for dynamic fracture of functional graded material is proposed in present work. The dynamic piecewise-exponential (DPE) model is extended to the meso-macro scale by the data mining method. With present model, the influence of meso-scale design parameters on macro-scale fracture behaviors can be investigated conveniently. Thus, the material design procedures and the fracture analysis of FGM are connected. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Fracture

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Research on the influence of ocean condition on the oscillation in parallel double-channel

Research on the influence of ocean condition on the oscillat...

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18th International Conference on Nuclear Engineering, ICONE18

作者： Zhou, Linglan Hong, Zhang Department of Physic Engineering Tsinghua University China National Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute Of China China

ISBN: (纸本)9780791849323

In this paper, the flow oscillation in the parallel multichannel system has been studied under ocean condition with RELAP5/MC. A double-channel boiling system model is built using the code RELAP5/MC and it is proved to be capable of simulating the instable phenomena. The influences of rolling, heaving and inclination on the flow oscillation have been analyzed. The result shows that the effect of lengthways rolling on the oscillation in parallel channels can be ignored. The influence of athwartships rolling and heaving reveals resonance effect, that is, when the rolling cycle is close to the oscillation cycle and half cycle respectively, the oscillation would happen earlier, when the difference between the cycles is large, the impact of rolling can also be ignored. The influence of inclination on the oscillation in parallel channels is feeble. © 2010 by ASME.

关键词： Oscillating flow

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Neutronic and Thermal-Hydraulic Analyses for an Additive Manufacturing reactor Core With SiC Matrix and TRISO Particle Fuel 29

Neutronic and Thermal-Hydraulic Analyses for an Additive Man...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Han, Wenbin Deng, Jian Lu, Qi Chen, Chong Xu, Youyou Zhang, Tao Huang, Shanfang 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)9780791886502

Additive manufacturing (AM) is a transformational digital manufacturing technology featured with rapidity, customizability, precision, and economy, which is fundamentally altering the way components are designed and manufactured. AM enables the freedom of design, and makes full use of complexity of geometry which comes for free . Applying AM technology to nuclear industry can yield advanced reactor designs with function and structure matched for the best thermal, hydraulic and mechanical performance. In this work, an AM-informed reactor core design with silicon carbide (SiC) matrix and tri-structural isotropic (TRISO) particle fuel is proposed and analyzed. The core is an integrated 3D-printing SiC bulk with helical cruciform coolant channels, and the UO2-bearing TRISO fuel particles are dispersed in the bulk. A multi-physics analysis framework for irregular geometry is developed to analyze and further optimize the reactor design. The TRISO particle positions are generated with discrete element method (DEM). The reactor Monte Carlo code (RMC) and the commercial CFD software STAR-CCM+ are used for the neutronic and thermal-hydraulic analyses, respectively. RMC simulates the neutron transport to predict the effective multiplication factor and power distribution. STAR-CCM+ calculates the flow and heat transfer in coolant channels and heat conduction in solid matrix with the power distribution as the heat source. Preliminary results show that the power peaking factor FQ decreases below 1.65, the heat transfer area increases by 30.3% and the fuel peaking temperature decreases by 25 K. The optimized AM-informed design enjoys better neutronic and thermal-hydraulic performance than those with regular geometry. © 2022 by ASME.

关键词： 3D printers

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The Software safety analysis based on sfta for reactor power regulating system in nuclear power plant 23

The Software safety analysis based on sfta for reactor power...

引用

23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Liu, Zhaohui Liao, Longtao Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design. Copyright © 2015 by JSME.

关键词： Nuclear power plants

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Intra Fast Mode Decision Algorithm and Hardware design for AVS2

Intra Fast Mode Decision Algorithm and Hardware Design for A...

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2021 International Conference on Signal Processing and Machine Learning, CONF-SPML 2021

作者： Xie, Hao Liu, Kaiyang Zhao, Yang Zheng, Xiao Qing, Xianguo Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China Peking University School of Software and Microelectronics Beijing China

ISBN: (纸本)9781665417341

AVS2 is a kind of video coding standard proposed by China, it adopts 33 intra prediction modes to improve coding performance, while the computational complexity has increased dramatically. In order to reduce the complexity of AVS2 intra mode decision and make the AVS2 hardware encoder meet real-time requirements, this paper proposes the AVS2 intra fast mode decision algorithm and hardware design. The experimental results show that the intra mode decision algorithm and hardware design proposed in this paper can meet the throughput requirement of 1920x1080@60fps at a clock frequency of 300MHz. By using Xilinx FPGA XC7K325T 900 on the Vivado HLS platform for synthesis, only 10% of the LUT, 5% of FF, 5% of BRAM, and 6% of DSP in FPGA resources are consumed to meet the design requirements. © 2021 IEEE.

关键词： Field programmable gate arrays (FPGA)

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The investigateion of damage and fracture behavior of particle reinforced composites at mesoscale under complex impact loads 14

The investigateion of damage and fracture behavior of partic...

引用

14th International Conference on Fracture, ICF 2017

作者： Bai, Xiaoming Guo, Licheng Shi, Kaikai Zheng, Liangang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China Department of Astronautic Science and Mechanics Harbin Institute of Technology Harbin150001 China

A mesoscale numerical model for the damage and fracture behavior of particle reinforced composites is proposed in present work. In this model, the digital image technique is used to reconstruct the real microstructure of composite material. The damping boundary condition is applied to absorb the artificial reflected stress wave during the simulation. Thus, the damage and fracture behavior of particle reinforced material under complex impact loads can be investigated. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Particle reinforced composites

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Research on trans critical characteristic on sliding pressure booting process of SCWR

Research on trans critical characteristic on sliding pressur...

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2013 21st International Conference on Nuclear Engineering, ICONE 2013

作者： Liu, Liang Zhou, Tao Song, Mingqiang Chen, Juan Luo, Feng Xia, Bangyang North China Electric Power University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791855836

The homogeneous flow model of trans critical transient calculation is achieved by the trans critical program of SCWR which is made by FORTRAN. The SCWR sliding pressure starting scheme is simulated by this program. So that the pressure raising stage characteristic of its booting process is analyzed. The calculation result shows that, in the pressure raising stage of booting process, the maximum cladding temperature gradually increases. Coolant temperature increases gradually along with the flow direction. The dry degree also increases along with the flow direction. The highest temperature of cladding does not exceed 650 °C limits. There is no major adverse effect on heat transfer of fuel rod in the dramatic changes in the physical properties of water. Copyright © 2013 by ASME.

关键词： Heat transfer

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Calculation strategy for the determination of plasticity correction factors 13th

Calculation strategy for the determination of plasticity cor...

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13th Asia-Pacific Symposium on Engineering Plasticity and its Applications, AEPA 2016

作者： Du, Juan Shao, Xue Jiao Kan, Qian Hua Zhang, Ying Fu, Xiao Long Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Southwest Jiaotong University Chengdu610031 China

ISBN: (纸本)9783035710243

This paper presents an investigation about the plasticity correction factor, Ke, proposed in the RCC-M code, for the correction of elastic stress range exceeds twice the yield stress which results from both mechanical and thermal loading. The plasticity correction factors are analyzed using a calculation strategy based on the comparison between linear elastic and elastoplastic analyses. The RCC-M code provides the codified expressions of Ke and material parameters for general materials of components in nuclear power plant, but there are no expressions and material parameters for new materials. Several elastoplastic models, geometries and loading types are considered in this study based on the general materials listed in RCC-M code to determine the calculation strategy of Ke for new material. Based on the results, the expressions and parameters of Ke for titanium alloy are discussed. It turns the proposed Ke factor of titanium alloy can yield to sufficiently conservative results. A verification calculation of general materials is performed and the conservative allowance is obtained. © 2017 Trans Tech Publications, Switzerland.

关键词： Plasticity

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