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

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Research on periodic test scheme of safety digital control system for nuclear power plant 27

Research on periodic test scheme of safety digital control s...

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

作者： Xie, Weibo Wu, Zhiqiang Zeng, Shan Li, Yutong Hu, Qingren Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The main trend in the development of the local power plant control system is to adopt digital safety DCS system with rather strong ability of self-diagnosis. Apart from the self-diagnosis, it is also necessary to carry out periodic test in certain frequency to check the potential fault within a system and to certify the completeness and effectiveness of reactor protection function. Regarding nowadays in-service or in-building power plant, there are some deficiency in the automation and simplication of the periodic test method, which still can be improved. Based on the research on the overall optimization of nuclear power plant safety DCS periodic test, an integrated and comprehensive periodic test method was put forward. This method, not only achieved a high degree of automation, supporting T2 online test, which increasing the system reliability, in order to save the cost of human and material during window phase in the overhaul life. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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Constructing Basic Probability Assignment Based on the Slope Correlation Degree 5

Constructing Basic Probability Assignment Based on the Slope...

引用

5th IEEE International Conference on Electronics technology, ICET 2022

作者： Chen, Jie Yang, Pengcheng Pu, Xiaofei Xie, Hao Wang, Jinxin Wu, Kunren Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9781665485081

DS evidence theory can be applied to deal with multi-sensor data and get reliable and accurate decision result of system's target or object due to its good measurement and reasonable description of system's uncertainty. In practical application field, DS evidence theory is achieved by two steps, one is the acquisition of the basic probability assignment function (BPA), and the other is the synthesis of multiple evidences that is obtained from multi-sensor data. The existing relevant literatures mainly solve the combination problem of conflict evidences, while lack of systematic research on the acquisition of BPA. In this paper, we focus on the discussion and innovation of the acquisition of BPA. BPA in DS evidence theory is the basic unit to describe the uncertainty of the system and evidences, which fundamentally determines its superiority with respect to probability theory and Bayesian reasoning. Getting a reasonable and effective BPA is the premise of the application of DS evidence theory. Traditional BPA acquisition methods have the limitations like absent priori knowledge and low acquisition accuracy. Aiming to solve the technical difficulties of traditional methods, we put forward a new BPA acquisition algorithm based on the slope correlative degree. The novel algorithm proposed in this paper successfully realizes the reliable descriptions of system's uncertainty, which is conducive to product the accurate BPAs and the precise fusion results. Experimental results and analyses demonstrate that the innovative algorithm can not only produce reasonable BPAs, which can better fit the sensors' monitoring data, but also expand the gap between the support degree of different propositions in BPA, which is more beneficial to the synthesis of evidences and the decision fusion of the system. Thus, the novel BPA acquisition algorithm has great application value and research significance. © 2022 IEEE.

关键词： Uncertainty analysis

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Numerical simulation of sealing behavior of C-ring for reactor Pressure Vessel 23

Numerical simulation of sealing behavior of C-ring for React...

引用

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

作者： Dong, Yuanyuan Luo, Ying Zhang, Liping Nuclear Power Institute of China Science and Technology Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9784888982566

C-ring, whose sealing behavior is the key to the sealing structure, is one of the main seal-rings for reactor Pressure Vessel. A finite element model with less simplification has been established using ANSYS APDL to simulate the sealing behavior of C-ring. The numerical results, including sealing behavior curve, specific pressure and springback value, have been compared with test results at 7.67%, 8.42% and 8.09% compression ratios respectively. It confirms the validity of this numerical simulation method. With the numerical method, the deformation and stress distribution and sealing behavior curve of C-ring can be obtained. The behavior curve can be divided into 5 sections. At 10% compression ratio, the peak contact stress is 501MPa, the contact width is 3.22mm, the total springback value is 0.460mm, the effective springback value is 0.332mm and the linear load is 788N/mm. As C-ring is composed of three parts, three models, i.e. spring-only, spring with inner lining and spring with two linings, have been simulated to study the effect of the contact states between each parts on sealing behavior. Copyright © 2015 by JSME.

关键词： Numerical models

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Improved core design of the supercritical water-cooled reactor CSR1000

Improved core design of the supercritical water-cooled react...

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2017 Transactions of the American Nuclear Society, ANS 2017

作者： Wang, Lianjie Lu, Di Yang, Ping Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Numerical methods research and fracture failure analysis for critical crack size 27

Numerical methods research and fracture failure analysis for...

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

作者： Shi, Kaikai Yuan, Yanli Chen, Jianguo Bai, Xiaoming Zheng, Liangang Lu, Xifeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Safety analysis of fracture mechanics is an important part of structural integrity assessment and critical crack size is one ofmain concerns for fracture failure. From the perspective of mathematics, the calculation of critical crack size is the solving and analysis for nonlinear equation f(x)=0. About the solution of nonlinear equation, there are the bisection method, the false position method, the simple iteration method, the newton method and the secant method. For the complexity of load imposed on nuclear power equipment, there may be the multi-value phenomenon of critical crack size in analyzing fracture failure problems. Considering that the corresponding mechanical states of the shallow crack and the deep crack are different when the fracture failure occurs, the crack toughness factor needs to be considered in the fracture failure analysis of shallow cracks. Connecting the bottom head of nuclear power plant main equipment, the pipeline large break transient of accident condition and the welding research council bulletin 175, the algorithm and the convergence of multi-value phenomenon of critical size are studied. Using the crack toughness factor, the fracture failure critical crack size is deeply analyzed and discussed. The algorithm and fracture failure theory in the present study can provide the technical support for the prolonging operation life projects of nuclear power plant main equipment in future. Copyright © 2018-2019 by JSME

关键词： Numerical methods

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The study on overall I&C system scheme for floating nuclear power plant ACP100S and ACP25S 10

The study on overall I&C system scheme for floating nuclear ...

引用

10th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2017

作者： Chen, Zhi Long-Tao, Liao Dan-Rong, Song You, Kai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781510851160

Floating nuclear power plant (FNPP) is built on the offshore platform and it is defined as a kind of Small and Modular reactor(SMR) due to its power level. The FNPP can provide power supply, fresh water and high temperature steam and other products. So it can be used as special energy sources, such as regional power supply, district heating and offshore oil exploitation. At present, two types of FNPP, named ACP100S and ACP25S, are developed by China national Nuclear Corporation (CNNC). In this paper, the overall technical parameters and overall technical scheme of ACP100S and ACP25S are introduced firstly. On the basis of these schemes, the different arrangements of nuclear power plant which are used for meeting the different energy demands are discussed. Due to the different configuration of nuclear power plant, the requirements for the Instrument and Control (I&C) system are different, so the paper discusses some overall configurations of I&C system under the several typical arrangements models, i.e., the single reactor and the single turbine, the double reactors and double turbines, and the double reactors and the single turbine. Furthermore, the characteristics of these overall I&C configurations are analyzed. © 2017 Westinghouse Electric Company LLC. All Rights Reserved.

关键词： Nuclear power plants

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The Optimization design of the reactor Coolant system Based on Optimus 23rd

The Optimization Design of the Reactor Coolant System Based ...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Yanli, Yuan Xianhui, Ye Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Structural dynamic analysis of reactor coolant systems is an important means of evaluating the safety of nuclear power plants under seismic loads. The seismic load distribution of the equipment can be obtained through the dynamic analysis of the reactor coolant system, and the stress of the equipment can be analyzed. In order to reduce the seismic stress of the equipment, effectively reducing the seismic load of the equipment is an important way. In this paper, aiming at the problem of excessive load at the outlet nozzle of the reactor pressure vessel, combined with the design characteristics of the reactor coolant system under the seismic, the optimal design of the reactor coolant system is carried out by the OPTIMUS [1] which is a kind of software developed by Noesis Solutions Company, by which the multidisciplinary analysis processes can be integrated for collaborative optimization design, and ANSYS [2], MATLAB and other software can be integrated to automate the simulation processes. Firstly, a nonlinear model for the dynamic analysis of the reactor coolant system is established, and then the dynamic response sensitivity analysis of the reactor coolant system is carried out, from which the design variables for subsequent optimization analysis are selected, and then the optimization analysis is processed with the force load at the outlet nozzle of the reactor pressure vessel minimized as the target function, and with the force load at the inlet nozzle of the reactor pressure vessel, the bending moment load at the inlet nozzle of the reactor pressure vessel and the bending moment load at the outlet nozzle of the reactor pressure vessel as the state variables. The results show that after the optimization, the load at the outlet nozzle of the reactor pressure vessel is effectively reduced. Relevant optimization analysis methods and analysis processes can be extended to other systems and equipment in the nuclear power plants. © 2023, The Author(s), under exclusive li

关键词： Nozzles

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INVESTIGATION ON THE APPLICATION OF MRF DAMPER BASED ON-OFF SEMI-ACTIVE VIBRATION CONTROL ON THE reactor COOLANT PUMP 30

INVESTIGATION ON THE APPLICATION OF MRF DAMPER BASED ON-OFF ...

引用

30th International Congress on Sound and Vibration, ICSV 2024

作者： Liao, Guojiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789090390581

The seismic design of the reactor coolant pump is important for the nuclear power plant. In this study, the magnetorheological fluid damper (MRF damper) is introduced to the supporting structure of reactor coolant pump, which is referred to HPR1000. MRF damper is a kind of intelligent damper, whose damping force can be controlled by the external magnetic current within several milliseconds according to the external excitation environment. In this study, considering both the controllable damping property of MRF damper and the complexity of the control strategy, the relative velocity/displacement based ON-OFF control algorithm is proposed to control the damping force of MRF damper. Numerical simulation demonstrates that the absolute acceleration and the relative displacement of the reactor coolant pump under seismic excitation are suppressed greatly with the help of MRF damper. The MRF damper is efficient in improving the seismic safety of the nuclear grade reactor coolant pump. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Damping

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An automatic optimization method for nuclear piping layout and mechanical analysis 27

An automatic optimization method for nuclear piping layout a...

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

作者： Bai, Xiaoming Lu, Xifeng Zhang, Rui Shi, Kaikai Zheng, Liangang Huang, Qian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The piping system in the nuclear power plant is extremely complex. The traditional trial method for the optimization is time-consuming and experience-dependent. Moreover, it rarely obtains the layout with best mechanical property using the trial method. In order to improve the traditional method, an automatic optimization method is proposed in present work. The genetic algorithm is employed during the mechanical analysis of piping system. In this method, both the piping layout and the piping support can be optimized concurrently. Moreover, based on the optimization method, a piping optimization theory assistant tool (POTATO) is developed. Some optimization instances in the nuclear power plant is shown in this paper. Comparing with the layout from the trial method, the one from present automatic method have more mechanical margin. The stress ratios at all the conditions are less than the limitation. Meanwhile, the time cost in present method is less than the trial method. Thus, present method is more efficient than the traditional method, and is significant for the piping layout in nuclear engineering. Copyright © 2018-2019 by JSME

关键词： Genetic algorithms

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The defect analysis technology study of reactor pressure vessel main bolt 27

The defect analysis technology study of reactor pressure ves...

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

作者： Zheng, Liangang Zhang, Rui Huang, Qian Bai, Xiaoming Yuan, Feng Lu, Xifeng Yuan, Yanli Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

The existence of non conformances and defects are inescapable during the manufacture and in service of reactor pressure vessel (RPV). After being analyzed, most of the defects have no influence on the capability and safety of the structure if the analysis results satisfy criteria. The defects including main bolts defects, the non conformances of cladding thickness and structure dimension, structure scratch, material inclusion and so on. The typical and serious defects is main bolts defects, including thread defect, stress relaxation of bolt and the extreme case of bolt break. The results of stress, fatigue and seal performance of reactor pressure vessel will be changed considering these defects. Based on the mechanics theory and numerical simulation technique, a kind of stress, fatigue and seal performance analysis method about the defects effects is studied in this paper. The influence to the reactor pressure vessel induced by several kinds of bolt break accident is analyzed using this analysis method. Copyright © 2019 by JSME

关键词： Fatigue of materials

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