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

共 626 条记录，以下是161-170 订阅

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The CFD calculation of a stratification break-up test of light gas in a closed vessel

The CFD calculation of a stratification break-up test of lig...

引用

2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Yazhe, Lu Yu, Zhang Xu, Ran Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784890471676

During the course of a hypothetical severe accident in a light water reactor, spray systems could be used in the containment in order to prevent overpressure in case of a steam line break and to enhance the gas mixing in case of the presence of hydrogen. Thus these two effects induced by spray might compete with each other having an influence on the hydrogen risk. Focusing on the dynamic effect of spray, the impact of spray on the breakup of a light gas layer in a closed vessel is simulated and the experimental test evaluated here is TOSQAN 113 test that belongs to the dynamic part of SARNET FP6 benchmarks. ANSYS FLUENT 14.5 is chosen for the simulation tool and spray droplets are tracked with using the Euler-Lagrange approach-Discrete Phase Model (DPM). The code validation and several sensitivity studies are done to sum up the basic method of numerical simulation of spray with CFD code and assess the influence of calculation settings and parameters, such as turbulence model and spray boundary condition. The code validation shows that the results of calculations acceptably agree with the experimental data and the sensitivity studies reveal that spray boundary condition is a rather important factor for the accuracy of simulation. Conclusion gained in from the present work can provide some reference of the numerical analysis of containment.

关键词： Boundary conditions

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A calibration method for raman distributed temperature sensor in nuclear power plant

A calibration method for raman distributed temperature senso...

引用

2019 Transactions of the American Nuclear Society, ANS 2019

作者： Xiang, Meiqiong Liu, Yanyang Qing, Xianguo Zhu, Jialiang Wu, Qian Zhu, Biwei Xu, Sijie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Maap5 simulation of pwr high-pressure core melting severe accident due to smallbreak loca

Maap5 simulation of pwr high-pressure core melting severe ac...

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2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Wu, Xiaoli Li, Wei Deng, Jian Deng, Chunrui Hou, Liqiang Zhang, Dan Yuan, Hongsheng Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780894487552

MAAP5 is capable of simulating the whole response of a PWR nuclear power plant under severe accident conditions. An analysis of the severe accident scenarios in a PWR by MAAP5 were carried out. It was assumed that the severe accident was initiated by an unusual smallbreak loss-of-coolant accident (SBLOCA) at the reactor pressure vessel (RPV) lower head and the failure of highpressure injection system (HPIS) in a PWR by using MAAP5 were carried out. Different break sizes with and without manual depressurization of the reactor coolant system (RCS) were studied. The results showed that without manually depressurization the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater work also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture took place. © 2018 American Nuclear Society. All rights reserved.

关键词： Loss of coolant accidents

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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network 4

Research on Hybrid Modeling Method Based on Mechanism Model ...

引用

4th Asia Conference on Information Engineering, ACIE 2024

作者： Chen, Jie Zheng, Yanqiu Chu, Qing Jiang, Guanfu Li, Jing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9798350371925

reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are not conductive to the transplantation and improvement of reactor system. In order to realize the rapid development of reactor control system, this paper carries out the mechanism modeling of the reactor system in Matlab/Simulink, and then introduce BP neural network to establish the hybrid model, which improves the modeling accuracy to support the design of reactor system. For small pressurized water reactor, we firstly take analytical modeling method to build the core mechanism model and its state space model through three conservation principles and the operating mechanism of core. Based on analyses, we can conclude that the parameter error and structural error of the mechanism model is caused by the assumption and simplification during modeling, and choose the parameter with the most influence (the feedback coefficient of the coolant temperature) as the corrected parameter in the BP neural-based hybrid modeling process. After that, according to the dual constant operation scheme of small pressurized water reactors, and the gradient descent method of BP network, the hybrid model of core is built and the parameter error and structure error of the mechanism modeling are corrected online. The simulation results show that the proposed hybrid modeling method can accurately simulate the reactor operate state. Thus, the proposed hybrid model is accurate and effective. © 2024 IEEE.

关键词： Pressurized water reactors

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Theoretical exploration of heat transfer characteristics of liquid metals

Theoretical exploration of heat transfer characteristics of ...

引用

2020 Transactions of the American Nuclear Society, ANS 2020

作者： Shen, Yaou Peng, Shinian Yan, Mingyu Zhang, Yu Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

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SMR built-in pressurizer operation transients analysis

SMR built-in pressurizer operation transients analysis

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2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Songwei, Li Yu, Liu Xi, Chen Zhongchun, Li Danrong, Song Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784890471676

Three primary design principles, meanwhile the design features, are carried through the R&D process of Chinese Small Module reactor (ACP100): integral layout, compact design and modularized configuration. As a substantial demonstration of the above design conceptions, an upper plenum built-in pressurizer is proposed for the pressure control system design for Chinese SMR. In this design, branches with large diameters connected to the primary loop (i.e. surge line, etc.) are eliminated, and this will significantly reduce the LOCA possibility due to the pressure boundary breach. This study mainly focuses on the feature of responses of the built-in pressure control system during the system transients, such as step load increase/decrease, ramp load increase/decrease and partial loss of electrical load. Initial status of the plant, assumptions, postulated transient condition and methodology of this analysis are described in the first section. Important thermal hydraulic parameters that can picture the transient characteristics of the coolant system are given in the second section. It shows that, during those transients, no reactor safety related signals or safety guard features are triggered or actived. It also shows that Nuclear/thermal power response rapidly and stably, and average coolant temperature and system pressure are welled controlled within the range of normal deviation of the set point values of safety related system. This study results get a good agreement with the operating experiment study of the build-in pressure control system.

关键词： Transient analysis

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Research on boundary disposal of pipes decoupling based on PEPs 27

Research on boundary disposal of pipes decoupling based on P...

引用

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

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

ISBN: (纸本)9784888983051

Generally, one of the remarkable characteristics is: compared with the main pipe, the sizes of the auxiliary piping are smaller, which makes the auxiliary piping satisfy the criteria to be decoupled from the main piping. So when the auxiliary pipes connected to the main piping are analyzed, there are two methods: one is that the auxiliary pipe is analyzed connected to the main piping, and the other is that the auxiliary pipe is analyzed not to be connected to the main pipe. Because the response spectrum at the location of the connection of the auxiliary pipes and the main piping is lacked, the previous choice is usually adopted. At present, the main software used for analyzing the stress of the pipes in nuclear power station is PEPS, but when the PEPS is adopted to analyze the small pipes, if the small pipes are analyzed connected to the main pipe, the boundaries of the thermal analysis and those of the seismic analysis are different, so the two kinds of analysis cannot be carried out at the only one model. That is, when the small pipe is analyzed connected to the main piping, two models are needed for thermal analysis and the seismic analysis, and then the results of the thermal analysis and the seismic analysis are distilled separately, and the results are combined according to determinate mode, and then the combined results are evaluated according to the criterion. Consequently, not only is the model complex, but also the combination of the seismic stress and the thermal stress is not convenient in the previous method. In the latter method, the auxiliary pipe is not connected to the main piping, accordingly, not only is the modal simple, but also the combination of the seismic stress and the thermal stress is convenient. The only problem which is needed to be solved is the seismic response spectrum and the displacements, and an approach to solve the problem is presented in the article. A response spectrum calculation method is obtained based on the single oscilla

关键词： Thermoanalysis

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Research on Thermal-Hydraulic Model and Characteristics of Lead Cooled Traveling Wave reactor 23rd

Research on Thermal-Hydraulic Model and Characteristics of L...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Zhang, Xue Yu, Hongxing Du, Sijia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

Traveling wave reactor (TWR) is an advanced nuclear power system, which can keep the total amount of fissionable nuclides constant during its lifetime through the transformation of fissionable nuclides in the reactor, so as to maintain the stability of the effective proliferation factor of the reactor core. As a coolant, lead has low neutron moderation ability and capture cross section. At the same time, it also has excellent thermal characteristics. Lead cooled traveling wave reactor is a reactor with high inherent safety and good application prospects. The research on reactor core mainly focuses on the physical calculation of the reactor, and the analysis of its thermal hydraulic model and characteristics is relatively scarce, which limits the design and safety analysis of the advanced reactor. Therefore, it is necessary to investigate the thermal hydraulic model of lead cooled traveling wave reactor. In this paper, a thermal hydraulic model suitable for the core of lead cooled traveling wave reactor is proposed. Considering the axial particularity of heat release from the core of traveling wave reactor, the model considers the axial thermal conductivity of fuel rods ignored by other thermal hydraulic models. Due to the strong thermal conductivity of core coolant, the thermal model also considers the heat transfer of fuel assemblies and the flow distribution of core flow. The research shows that the thermal hydraulic model is suitable for the analysis of thermal hydraulic characteristics of lead cooled traveling wave reactor, the thermal hydraulic characteristics are studied by using the thermal hydraulic model. © 2023, The Author(s).

关键词： Thermal conductivity

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ANALYSIS OF SMR reactor COOLANT system IN APROS 31

ANALYSIS OF SMR REACTOR COOLANT SYSTEM IN APROS

引用

2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Zhu, Ye Xingbo, Wang Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780791888247

The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic of these reactors differs widely from the tradition nuclear power plant. APROS Nuclear simulation and analysis software, with the Comprehensive plant model covering reactor island, turbine island, electrical and automation systems, has been successfully used in a series of major nuclear power plant projects, including power upgrade, modernization, safety improvement, and new plant projects. A reactor coolant system model of SMR that contains both process and control elements is developed and simulated using APROS. Steady states analysis is conducted under different power load and both force/natural cycle. Analysis of Transient condition including increasing/decreasing power load and force/natural circulation switching is also conducted to acquire dynamic thermal hydrate parameter. Reference and guidance can be provided for system optimization and safety analysis. Copyright © 2024 by ASME.

关键词： Nuclear power plants

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Multiple branches coupling property and design of PRHRs

Multiple branches coupling property and design of PRHRs

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AISTech 2018 Iron and Steel technology Conference and Exposition

作者： Zhuohua, Zhang Xu, Ran Feng, Li Lin, Xian Ke, Zhou Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781935117728

Nuclear reactor applied on ship or floating platform recommend miniaturization and optimization of PRHRs. Due to much larger volume and heat transfer area in PRHRs heat exchanger, miniaturization and design of PRHRs need consider many constraints such as volume and systems configuration in containment. In normal reactor design such constraints are not so rigorous, but in reactor design for floating platform it takes significant difficulties. A multiple branches coupling system is constructed in the paper which describes the actual relationship between PRHRs and other systems or equipment, and by combining temperature distribution in heat exchanger and conservation equations of coupling system, the design parameters set satisfied design requirement can be calculated and optimization of design can be completed by programming solving method. © 2018 by AIST.

关键词： Heat transfer

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