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

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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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Numerical research on rotating speed influence and flow state distribution of water-lubricated thrust bearing 23

Numerical research on rotating speed influence and flow stat...

引用

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

作者： Deng, Xiao Deng, Li-Ping Huang, Wei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784888982566

Water-lubricated thrust bearing is one of the key parts in canned motor pump, for example, the RCP in AP1000, and it functions to balance axial loads. A calculation model which can handle all flow state lubrication performance for water-lubricated thrust bearing has been presented. The model first includes laminar and turbulent Reynolds' equation. Then to get continuous viscosity coefficients cross critical Reynolds number, a transition zone which ranges based on engineering experience is put up, through which Hermite interpolation is employed. The model is numerically solved in finite difference method with uniform grids. To accelerate the calculation process, multigrid method and line relaxation is adopted within the iteration procedure. A medium sized water-lubricated tilting pad thrust bearing is exampled to verify the calculation model. Results suggest that as rotating speed enlarges, lubrication state distribution of the thrust bearing gradually tends to turbulent lubrication from the intersection corner of pad outer diameter and pad inlet to the opposite, minimum water film thickness increases approximately linearly, maximum water film pressure has little change, meanwhile the friction power grows nearly in exponential law which could result in bad effect by yielding much more heat. Copyright © 2015 by JSME.

关键词： Iterative methods

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Fatigue crack growth model for edge sliding mode 14

Fatigue crack growth model for edge sliding mode

引用

14th International Conference on Fracture, ICF 2017

作者： Shi, K.K. Zheng, B. Bai, X.M. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The inherent law between fatigue behaviours of shear-type representative volume element and mode-II fatigue crack growth is found in the range of cycle plastic zone near the crack tip. Prediction models for mode-II fatigue crack growth rate are then proposed by utilizing shear-type low cycle fatigue properties, fatigue failure criterions (plastic strain energy and linear damage accumulation), and effective cycle stress-strain field ahead of the crack tip. Experimental data of two aluminium alloys, 2024-T351 and 7075-T6, are used for models verification. Good agreement between experimental and theoretical results is obtained. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Fatigue crack propagation

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SMR built-in pressurizer operation transients analysis

SMR built-in pressurizer operation transients analysis

引用

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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A study of heat transfer calculation method of reactor vessel metallic insulation 22

引用

2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Dapeng, Yan Luo, Ying Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791845899

Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter "view factor", and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster. Copyright © 2014 by ASME.

关键词： Heat radiation

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Neutronic feasibility analysis of fully ceramic microencapsulated fuel for commercial PWRs

Neutronic feasibility analysis of fully ceramic microencapsu...

引用

Top Fuel 2016: LWR Fuels with Enhanced Safety and Performance

作者： Mancang, Li Dan, Wang Yingrui, Yu Dong, Qin Wenjie, Li Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9780894487309

The fully ceramic microencapsulated (FCM) fuel is based on tristructural isotropic (TRISO) coated particles embedded in silicon carbide (SiC) matrix, which is among the accident tolerant fuel concept options. With the strength of high fission product retention, the FCM fuel provides potential superior safety performance. Compared to the conventional standard UO2 fuel, the deficiency of FCM fuel is its considerable lower fissile inventory, which results in a reduced reactor power or/and life. Supposing the FCM fuel is used as direct replacement of the fuel in commercial PWRs, the spectrum is softened due to the usage of SiC as matrix, and the moderator temperature coefficient is possible to be positive at the beginning of life, leading to high risk during operation. A neutronic feasibility study of FCM fuel utilized in commercial PWRs is performed in the paper. The fuel enrichment and undermoderated lattice are evaluated by analyzing the typical TRISO particle with UO2 kernels which is widely used in high temperature gas reactors and that with enlarged UN fuel kernels. It was found that under the mandatory constraint of low enriched uranium, the FCM fuel with UO2 kernels, although relatively mature, is only capable of small reactor designs, and the assembly lattice needs to be modified. The FCM fuel with UN kernels, when directly loading in the commercial PWRs, can yield comparable reactor power and life;however, a higher fuel enrichment and the fuel design and fabrication need to be optimized.

关键词： Silicon carbide

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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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Study on Uniaxial Tension and Cyclic Deformation Behavior of TA16 at Room Temperature and High Temperature

Study on Uniaxial Tension and Cyclic Deformation Behavior of...

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作者： Shao, Xuejiao Juan, Du kuang, Linyuan Lu, Yuechuan Lu, Jiang Yu, Mingda Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

TA16 is a new type of titanium alloy material, which is widely used in pipeline systems of aerospace, ship and nuclear reactor with high temperature and high pressure. In this paper, the basic mechanical properties and cyclic deformation characteristics of TA16 were obtained by uniaxial tensile test and cyclic loading test at room temperature and high temperature. The Chaboche constitutive model can reasonably describe the cyclic plastic cumulative effect of materials and is provided in ANSYS which is a commercial finite element analysis software. The constitutive model of Chaboche presents a very important evolution criterion of follow-up hardening for ratchet simulation. Based on steady cyclic stress-strain curves which is got by strain control at different temperatures, the material parameters of Chaboche model are obtained by least multiplication fitting for TA16 at room and high temperatures. The constitutive model can used in elastic and elastic finite element analysis of TA16 materials. The experiment results show that TA16 exhibits slight cyclic softening under symmetrical strain cycles, and obvious ratcheting behavior under asymmetric stress cycles. When ratcheting strain of the structure reaches a certain saturation state, plastic accumulation will lead to accidental damage of the structure and lead to serious consequences. The stable cyclic stress-strain curves of TA16 at room temperature and high temperature are simulated by Chaboche model. The simulation results are in good agreement with the experimental results, which provides the basic data for the finite element simulation of cyclic deformation behavior of TA16 titanium alloy structure. © Springer Nature Switzerland AG 2020.

关键词： Constitutive models

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Prospects for Next Generation Nuclear Power system 23rd

Prospects for Next Generation Nuclear Power System

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Yu, Hongxing Xia, Bangyang Li, Wenjie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811988981

systematically research has been carried out on the demand of economic development for low-carbon clean energy and the opportunities faced by the development of nuclear power. A new generation of nuclear power technical index of "Inherent safety, Economics competitiveness, Environment-friendly, and Sustainability" has been formed, which provides guidance for the innovation and development of nuclear power technology. Considering the technical potential of various advanced reactor including the fourth-generation nuclear power system, inherently safe multi-purposed lead-cooled traveling wave reactor is proposed to solve the long-standing problems that have confronted nuclear power. The advanced reactor has attracted much attention because of their unique advantages in fuel supply issue, inherent safety, and economic efficiency. In this study, the main features of the advanced reactors are reviewed and current challenges in their development are pointed out. The advanced reactor obviates the fuel supply issues, including the nuclide enrichment, while making abundant natural uranium usable nuclear fuel. Each of these novel features is technologically separable, so that these new reactor design concepts can be also applied piecewise to enhance prospects in many different utilization circumstances. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear energy

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Computation of fluid thrust force from the pressure release tank of nuclear power plant pressurizer 17

Computation of fluid thrust force from the pressure release ...

引用

17th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2017

作者： Xiong, Fu-Rui Jiang, Nai-Bin Wu, Ge Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

Pressure release tank is a key component of nuclear power plant safeguard. Evaluating the fluid thrust force from high pressure and high temperature flow jet under postulated accidents is an essential task for nuclear system design and safety estimation. In this work, flow jet impingement from pressure release tank to nearby building roof is computed via a two-stage strategy. First, the RELAP5 code is applied to estimate thermal-hydraulic conditions at pressure release tank outlet. The thermal-hydraulic computational model includes key components of pressure release tank and its connection to the pressurizer. Based upon the thermal-hydraulic estimation within the pressure release tank, detailed computational fluid dynamics (CFD) simulation is performed with full considerations of how flow thrust is applied on nearby structures. The two-stage computational strategy provides a practical paradigm for fluid thrust force estimation. © 2016 Association for Computing Machinery Inc. All Rights Reserved.

关键词： Computational fluid dynamics

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