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

作者： Li, Xing Qi, Peiyao Tan, Sichao Mi, Zhengpeng Li, Dongyang Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin Heilongjiang China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

The transient fluctuation of flow rate in a nuclear reactor is a common phenomenon in floating reactors and possible accident scenarios, the response for transient flow determined the damage threshold of fuel assemblies. Friction characteristics and velocity distribution of low-frequency pulsatile flows in a rod bundle have been studied. The experiments were conducted under the time-averaged Reynolds number 1000 Copyright © 2018-2019 by JSME

关键词： Pressure drop

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Pipe Routing of reactor Based on Adaptive A* Algorithm Combined With Genetic Algorithm

Pipe Routing of Reactor Based on Adaptive A* Algorithm Combi...

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Chinese Automation Congress (CAC)

作者： Weihang Lv Na Qin Xinyu Zhao Peng Yuan Jie Huang School of Electrical Engineering Southwest Jiaotong University Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (数字)9781728176871

ISBN: (纸本)9781728176888

To solve the complex 3-D space pipe routing problem, an algorithm framework combining genetic algorithm and adaptive A * algorithm (AAGA) is proposed. The initial population of the genetic algorithm is generated using the adaptive A * algorithm combined with the concept of intermediate join points, this method is also used in the generation of sub-paths in the crossover and mutation operations of genetic algorithms. On the one hand, this method improves the quality of the initial population and speeds up the optimization of genetic algorithm, on the other hand, it solves the problem that the genetic algorithm combined with the traditional A * algorithm (AGA) is prone to redundant inflection points. Finally, on the premise of ensuring pipeline connectivity, pipeline length and number of inflection points, the constraints of pipeline center of gravity and minimum bending radius are added to carry out simulation experiments. Simulation results show that the AAGA has better performance than adaptive A * algorithm and AGA.

关键词： Pipelines Genetic algorithms Layout Statistics Sociology Biological cells Routing

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High-accuracy time synchronization method in highly reliable fiber-channel communication system 27

High-accuracy time synchronization method in highly reliable...

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

作者： Chen, Qi Jiang, Wei Tu, Xiaodong Yu, Bo Liu, Mingxing Zhang, Xu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China School of Information and Communication Engineering University of Electronic Science and Technology of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Accompanied with strict requirements on accuracy and determinacy in nuclear power industrial safety Digital Control system (DCS), time-triggered technology has been used in Ethernet widely, while another network protocol, Fiber-Channel (FC) protocol which is popular in avionic electronic systems with hyper reliability in port to port communication stands still, even though the optical fiber cable is quite adopted by other protocols like Ethernet. One important reason is the lack of the capability to handling time sensitive tasks in FC protocol itself. In order to bring Fiber-Channel protocol into real-time applications, like safety DCS for nuclear industry, with the situation that there are pretty few researches on the time synchronization in FC protocol, this paper makes fundamental research on the solution to achieve high-accuracy time synchronization in FC protocol networks, which has been proved a vital important step to realize real time communication according to Time-Triggered Ethernet (TTE)(H. Kopetz al, 2005). By analyzing current Fiber-Channel network time communication strategies and demo-boards testing, this paper designs a method to achieve the same level or better synchronization accuracy in FC network with that in TTE, which figures out a new direction in this blank field of time synchronization in FC protocol. Furthermore, this method only asks for few resources and slight modifications on current communications for future development and improvement. Copyright © 2018-2019 by JSME

关键词： Accident prevention

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Hydrogen risk analysis of a large pressurized water reactor using hydragon code 27

Hydrogen risk analysis of a large pressurized water reactor ...

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

作者： Zhang, Xiaolong Yu, Jiyang Huang, Tao Jiang, Guangming Zou, Zhiqiang Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

This paper presents the hydrogen risk analysis of a large Pressurized Water reactor (PWR) undergoing a severe accident using the HYDRAGON code. The code has been developed by the Department of Engineering Physics, Tsinghua University, as a hydrogen risk analysis software specialized for containment building of nuclear reactor power plants. The chemical interaction between zirconium cladding of the fuel rods and overheated coolant may release hydrogen when severe accidents occur in PWRs. The hydrogen mixed with air existing in containment may jeopardize the integrity of containment building if the flammable mixture explodes. In order to mitigate the hydrogen risk, many nuclear power plants install the Passive Autocatalytic Recombiners (PARs) that allow hydrogen and oxygen react chemically at low temperature without any sights of flame. Recombiners remove hydrogen and thereby prevent explosions. This work simulated the gas diffusion behaviors during the accident in the containment of a large PWR. The simulations obtained spatial concentration distribution of various gas species including hydrogen, oxygen, steam and nitrogen and several key parameters that are required to complete hydrogen burning and explosion investigation. Sensitivity analyses of recombiners' quantity and positions and hydrogen discharge intensity provides the reference for hydrogen mitigation measure design in the containment. The results demonstrate that the number of recombiners rather than positions has stronger influences on hydrogen mitigation. Since the installation position reflects less sensitivity, the recombiners could be arranged as higher as possible if necessary. Copyright © 2018-2019 by JSME

关键词： Hydrogen

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Research about the influence of passive residual heat removal system on SGTR accident 27

Research about the influence of passive residual heat remova...

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

作者： Wu, Yirui Shan, Jianqiang Xu, Tao Ma, Haifu Yang, Junyun Yu, Hong Wang, Junlong School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an Shaanxi Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan Province China

ISBN: (纸本)9784888983051

Since the passive safety system turns out to be a desirable method to achieve simplification and increase the reliability of the performance of essential safety functions, passive technology with inherent safety features has been generally applied to the design of advanced nuclear power plant. The PWR (pressurized water reactor) NPP (nuclear power plant) in this paper is a kind of mature pressurized water plants in China, and is proposed to be improved with some passive systems to enhance safety performance. The proposed and adapted passive safety systems for this PWR power plant in this paper contain: (1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);(5) the PRHRS (passive residual heat removal system), which is installed on the reactor coolant system (RCS) loop. In this research, the behaviors of passive safety systems are studied and transient characteristics of the primary and secondary side of the reactor during the Steam Generator Tube Rupture (SGTR) accident are investigated using the best-estimate transient simulation code Relap5/MOD3.3. Results reported in this paper indicate that making use of passive safety systems especially the PRHRS could ensure reactor safety during the SGTR accident. Copyright © 2018-2019 by JSME

关键词： Pressurized water reactors

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Analysis and verification of burnup algorithm in reactor 27

Analysis and verification of burnup algorithm in reactor

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

作者： Hu, Yuying Luo, Qi Yao, Dong Yu, Yingrui Liao, Hongkuan Zhou, Bingyan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Burnup algorithm analysis is an important part of the physical calculation in reactor. The computing accuracy of burnup algorithms is directly related to the correctness of the whole physical calculation. Due to the difference from the rates and lives of varieties of nuclides, burnup matrices are always sparse rigid exponential ones. At present, methods to solve exponential matrices are divided into two types, numerical methods and analytical methods. Besides, numerical methods conclude series methods, ordinary differential equation methods, polynomial methods, matrix decomposition methods and so on. In this study, some common methods to solve burnup matrices are introduced, such as Chebyshev Rational Approximation Method(CRAM), Krylov Subspace Method, Pade Approximation Method and Transmutation Trajectory Analysis(TTA). Analysis and verification for microscopic point-depletion calculation are carried out with two typical burnup chains. One of the burnup chains for verification is a decay chain of neptunium-237(237Np), which contains 13 nuclides. The other chain is an actinide nuclide reaction chain, including 25 nuclides from decay and neutron capture and 9 nuclides from fission. Because there is no close annual chain in the burnup chains used in verification, the results of using Transmutation Trajectory Analysis were thought to be accurate, and it has been used as the benchmark for comparative analysis of other methods. The final result shows, Chebyshev Rational Approximation Method has advantages in calculation accuracy and simplicity, and it is suitable to be used in large burnup calculations. Copyright © 2018-2019 by JSME

关键词： Isotopes

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Comparison of SGTR accident with two different passive residual heat removal systems 27

Comparison of SGTR accident with two different passive resid...

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

作者： Wu, Yirui Shan, Jianqiang Xu, Tao Ma, Haifu Yang, Junyun Yu, Hong Wang, Junlong Deng, Yaxin School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an Shaanxi Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan Province China

ISBN: (纸本)9784888983051

Since the passive safety system turns out to be a desirable method to achieve simplification and increase the reliability of the performance of essential safety functions, passive technology has been generally applied to the design of advanced nuclear power plant. The traditional generation II+ nuclearpower plant is proposed to be improved with some passive systems to enhance safety performance. There are two series of passive safety systems for this kind of nuclear power plant in this paper, one of them contains: (1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);and (5) the PRHRS (passive residual heat removal system), which is installed on the reactor coolant system (RCS) loop;the other contains:(1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);(5) the PEFS(passive emergency feed water system), which is installed on the secondary side of SGs. In this research, the transient characteristics of SGTR accident with the two different passive residual heat removal systems which are PRHRS and PEFS is investigated using the best-estimate transient simulation code Relap5/MOD3.3. The calculation results show that both of PRHRS and PEFS could remove the decay heat from the primary loop efficiently to ensure the reactor safety and reduce the radioactive leakage to mitigate the accident impact during the SGTR accident. Due to the significant cooling effect of PRHRS and PEFS, the water level of broken SG would not be overfilled in both conditions, which is different with the SGTR accident of traditionalgeneration II+nuclear power plant. Compared with the PEFS, the PRHRS performed better in cooling efficiency, long-term cooling performance, reducing radioactive leakage capacity and mitigating accident effect. Copyright © 2018-2

关键词： Nuclear power plants

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Application of feedforward predictive control in DC furnace coordination system

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Journal of Physics: Conference Series 2021年第1期1780卷

作者： Zhiguang DENG HE Zhengxi ZHU Biwei ZHU Jialiang LV Xin WU Qian National Key Labortory of Science and Technology on Reactor System Design Technology Chengdu610213China

In order to solve the control difficulties such as large dynamic delay, serious coupling between machine and furnace and strong nonlinear, the feedforward and feedback predictive control is proposed to control the coordinated system on the basis of multi-model predictive control, and the static feedforward part designs the feed line feedforward of coal quantity against coal quality disturbance on the basis of the accurate balance of energy. Thus, the reference total fuel quantity under the target load instruction is determined, and the dynamic part fully considers the dynamic compensation of the unit energy storage, gives a certain overshoot fuel quantity to the boiler regulation, speeds up its response, and ensures the rapidity of the unit. The real-time simulation results of virtual DPU (Distributed Processing Unit) show that the designed feedforward-feedback predictive control can quickly respond to load change, improve the resistance to fuel disturbance, reduce the fluctuation of main steam pressure, and better ensure the safe, economic and stable operation of the coordination system.

关键词：

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Assessment of Coupled Effect of Steam Pipes on Seismic Analysis of reactor Coolant system

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IOP Conference Series: Earth and Environmental science 2021年第4期719卷

作者： Lijuan Li Zhenyu Liu Hui Li Xianhui Ye Honglei Ai National Key Laboratory of Science and Technology on Reactor System Design Technology Chengdu Sichuan China

reactor coolant system is connected with auxiliary pipes, surge pipes and steam pipes. In order to facilitate the dynamic analysis of reactor coolant system, it is necessary to decouple the pipes connected with the reactor coolant system. Although some international standards stipulating requirements for the decoupling of subsystems from the main system, in order to further explore the influence of decoupling steam pipes, the seismic response of reactor coolant system with or without steam pipes is simulated by using ANSYS program. Firstly, the modal analysis of reactor coolant system is carried out to analysis the influence of steam pipes on the resonant frequency of reactor system. Then, the standardized seismic dynamic analysis of reactor system with or without steam pipes is carried out. The nozzle load (inlet and outlet nozzle load of main device, such as reactor pressure vessel, steam generator, pump) and support load (of main device such as reactor pressure vessel, steam generator, pump) are output and compared. According to nozzle load and support load, it can be seen that the steam pipes have little influence on the seismic analysis of reactor coolant system, i.e. steam pipes can be decoupled in the seismic analysis of reactor coolant system.

关键词：

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The Application of Adaptive A-star Algorithm in Layout of Spatial Pipeline

The Application of Adaptive A-star Algorithm in Layout of Sp...

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Data Driven Control and Learning systems (DDCLS)

作者： Xinyu Zhao Na Qin Jie Huang Yongchen Miao the Institute of Systems Science and Technology Southwest Jiaotong University Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Harbin Shipping School Harbin China

ISBN: (数字)9781728159225

ISBN: (纸本)9781728159232

In order to settle the complex problem of layout of spatial pipeline, the diameter of pipeline is taken as the distance between adjacent scattered points, and the spatial layout including obstacles is approximated as a three-dimensional diagram of scattered points. A-star algorithm is used to search an optimal pipelines path from the three-dimensional diagram of scattered points. However, when traditional A-star algorithm addresses pre-selected points that are the same cost, it selects pre-selected points randomly. In view of the disadvantages, an adaptive A-star algorithm is proposed based on the traditional A-star algorithm, in this paper. The adaptive A-star algorithm selects more reasonable pre-selected point according to the surrounding environment and situation of pre-selected point , so as to achieve the requirements of the shortest pipeline and the fewest inflection point. Finally, the experimental simulation shows that the adaptive A-star algorithm has better performance than the traditional A-star algorithm.

关键词： Pipelines Path planning Layout Robots Heuristic algorithms Robot kinematics Optimization

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