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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Lei, Zhong Deng, Jian Li, Wei Wu, Xiaoli Deng, Chunrui Nuclear Power Institute of China Chengdu Sichuan China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791883785

Core melting and molten migration behavior are hot and difficult issues in the field of nuclear reactor severe accident research. The Moving Particle Semi-implicit (MPS) meshless method has potential to simulate free-surface and multiphase flows. In this study, the MPS method was utilized to simulate the melting process of UO2-Zr rod-type fuel elements. The models of heat conduction with phase change, simplified UO2-Zr eutectic reaction, viscous flow and surface tension were implemented with the framework of standard MPS method. Then, the improved MPS code was used to simulate and analyze the process of high-temperature melting and characteristics of molten migration and solidification in the coolant channel, aiming at revealing the severe accidents for light water reactors (LWR), particularly the early core damage. The results showed that compared with the case of higher initial temperature, when the initial temperature of molten UO2 is lower, more molten UO2 will solidify on the surface of rod cluster, and the blockage of upper flow channel caused by molten UO2 is more serious. In addition, this study also demonstrated the potential of the MPS method for the study of complicated severe accident phenomena in not only traditional LWR but also advanced nuclear reactors in the future. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Solidification

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The statistical analyses on the bubble contact diameters in the vertical narrow channel

The statistical analyses on the bubble contact diameters in ...

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Lu, Qi Liu, Yu Liu, Wei Chen, Xi He, Hangxing Zhou, Linglan Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China Nuclear Power Institute of China Chengdu610041 China

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Prediction of the critical heat flux in subcooled flow boiling in round tube using an improved mechanistic model

Prediction of the critical heat flux in subcooled flow boili...

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18th International Conference on Nuclear Engineering, ICONE18

作者： Na, Yu Yu, Zhang National Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu Sichuan 610041 China

ISBN: (纸本)9780791849323

Among the existing CHF models, the bubble crowding model and the liquid sublayer dryout model have been well accepted for subcooled flow boiling. But both of the two models couldn't give explanation about some details in the boiling crisis phenomenon according to photographic result. The aim of the present paper is to provide an improved synthesized model containing the characteristic of the above two models and then to give a comprehensive explanation about CHF. In the present model, the conservation equations of mass and energy are solved to derive the CHF formula. The length and velocity of the vapor blanket and the thickness of the liquid sublayer are needed. The quality and void fraction in bubble region and the core region are calculated by a homogeneous assumption. The vapor blanket length is thought to be equal to the Helmholtz wavelength and it is obtained from several parameters in the bubble region. The velocity of the vapor blanket is connected to the flow velocity of the bubble layer. The thickness of the sublayer is determined by a force balance on the vapor blanket, which is also related to the condition of the bubble region. About 1100 experimental points have been selected to verify the proposed model. Comparison between the predictions by the proposed model and the experimental result shows a good agreement that more than 90% of these data are predicted within ±20%. © 2010 by ASME.

关键词： Void fraction

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REVIEW OF RISK-INFORMED SAFETY design CONCEPT, METHOD AND APPLICATION 30

REVIEW OF RISK-INFORMED SAFETY DESIGN CONCEPT, METHOD AND AP...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Gong, Lanxin Zhong, Mingjun Peng, Changhong School of Nuclear Science and Technology University of Science and Technology of China Hefei China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Since the NRC issued the PSA policy in 1995, PSA containing risk information has been widely used in supervision and design. Subsequently, the concept of "risk-informed" has gained more and more recognition, and its application has expanded from the initial equipment maintenance to the current design. By analyzing the regulations, standards and other guidance documents of various countries and institutions on safety design, we reviewed the history of risk-informed safety design concepts, methods and their applications, especially the progress of LMP projects in the United States in recent years. We put forward prospects for the development and application of risk-informed concepts and methods in new reactors in the future. © 2023 by JSME.

关键词： design

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KYLIN-V2.0 code calculation ability verification based on VERA benchmark

KYLIN-V2.0 code calculation ability verification based on VE...

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2020 International Conference on Physics of reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020

作者： Wang, Dongyong Yu, Yingrui Peng, Xingjie Wang, Chenlin Liu, Kun Chai, Xiaoming Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781713827245

Virtual Environmental for reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance self-shielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference. © The Authors, published by EDP sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://***/licenses/by/4.0/).

关键词： Control rods

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Natural convective heat transfer from a heated slender vertical tube in a cylindrical tank 16

Natural convective heat transfer from a heated slender verti...

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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Xian, Lin Jiang, Guangming Yu, Hongxing Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781510811843

Natural convective heat transfer in enclosures is widely researched in extensive range of engineering application such as passive residual heat removal system (PRHRS) in nuclear plant, solar collectors and cooling of electronic equipment. After Fukushima incident, the application and researching of PRHRS have been gaining more and more attention. It still has no proper calculating tools to analyze natural convective heat transfer with large L/D ratio and high Rayleigh number in PRHRS. In this paper, based on some former natural convective heat transfer experiments, numerical simulations and scale analysis, the single phase natural convective heat transfer correlation from the vertical tube have been developed. Transient single phase natural convection heat transfer along vertical tube in a cylindrical enclosure has been investigated for water with widely used computational fluid dynamic (CFD) engineering program ANSYS CFX-12. The influence brought by different geometry of tube and mesh scales has been evaluated by the results of former experiments. The ratio of tube length to diameter is a dimensionless factor affects heat transfer rate. The evaluation of thermal stratification to natural convective heat transfer in enclosure has been studied. Numerical simulation covers large L/D and high Rayleigh number natural convection: 108RaL14. Based on the scale analysis of natural convection cylindrical boundary equations the new dimensionless group has been put forward contains L/D factor. The new natural convective heat transfer along the vertical single tube model has been build according to the former experiments1 results. After combining numerical simulation results, the new model covers the higher Rayleigh number range of natural convection.

关键词： Natural convection

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Research on analysis and modeling methods for lbloca in nuclear power plants based on autonomous loca analysis platform arsac

Research on analysis and modeling methods for lbloca in nucl...

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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Zhou, Jiayue Wu, Dan Ding, Shuhua Jiang, Guangming Nuclear Power Institute of China Chengdu Sichuan China Science and Technology on Reactor System Design Technology Laboratory Chengdu Sichuan China

ISBN: (纸本)9780791883785

In order to meet the demand of continuous innovation of technologies and the general trend of autonomous nuclear power plants design and export of nuclear power plants, it is necessary to develop an autonomous LOCA analysis platform and corresponding analysis methods for the most complex design basis accidents. In this paper, the characteristics of LOCA analysis platform ARSAC, designed by Nuclear Power Institute of China, and the code ARSAC-K which meets the requirements of the US Federal Code 10 CFR 50.46 Appendix K model are introduced as well as a set of LOCA analysis methods and modeling methods. Based on the international advanced LOCA analysis code development concept, the code ARSAC has made new breakthroughs in matrix algorithms, key thermal hydraulic models and so on. Validation work has also been carried out in-depth. A set of advanced LOCA analysis methods has been developed using code ARSAC-K and advanced power plant parameter sampling methods. Analysis on LBLOCA of nuclear power plants with code ARSAC-K was performed, and the impact of different modeling methods on the LOCA analysis results was studied. To ensure the rationality and conservativeness of the analysis results, a set of reasonable and conservative modeling methods is fixed on the basis of a large number of sensitivity analyses for subsequent analysis and calculation. In the future, a lot of optimization work will be done to improve the LOCA code and corresponding methods. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear power plants

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Preliminary study on the radiation heat removal capacity in SMR core under high temperature condition

Preliminary study on the radiation heat removal capacity in ...

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

作者： Li, Zhongchun Ding, Shuhua Li, Songwei Deng, Jian Nulcear Power Institute of China Chengdu610041 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9784890471676

The advanced fuel design, such as accident tolerant fuel (ATF) was design concept to withstand the high temperatures for extreme condition. When it was equipped in reactor core, the heat removal by radiation would be quite considerable. The radiation heat removal in a simplified SMR core was studied in the present study. The calculation was assumed to be 2D and at steady state. The calculation models were established from fuel assembly to fuel rod. The radiation heat flux was calculated based on the heat balance model when the limitation temperature in the core was reached. The results showed that the radiation heat flux contribute large proportion when the limitation temperature on fuel clad was high. The radiation heat flux was increased with fuel clad temperature limitation value. The higher the emissivity was, the more radiation heat flux removed.

关键词： Temperature

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Large eddy simulation of turbulent flow in the rod bundle with different spacer grids 23

Large eddy simulation of turbulent flow in the rod bundle wi...

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

作者： Wei, Zonglan Yu, Zhang Liu, Songtao Science and Technology on Reactor System Design Technology Laboratory NPIC Chengdu Sichuan China Nuclear Power Design and Research Sub-Institute NPIC Chengdu Sichuan China

ISBN: (纸本)9784888982566

Large eddy simulation(LES) of the swirling flow in the rod bundle subchannels with spacer grids are presented. According to the rod bundle flow benchmark experiment, the 5×5 rod bundle with a split-type spacer grid is simulated by this transient simulation with a fine hybrid mesh. Mean velocity components, RMS (root mean square) velocity components and turbulent kinetic energy profiles at different locations downstream the mixing vanes are compared between the experiment and simulation results. LES results can reproduce the mixing turbulent flow velocity field well. We also have studied the influence of the mixing vanes' bending degree on the subchannel flow structures by using LES to simulate the 3×3 rod bundle with the typical PWR spacer grid. The mixing vanes' bending degree is modified to 35°and 40°based on the original spacer grid, whose mixing vanes' bending degree is 30°. The larger bending degree produces a stronger mean coolant mixing effect, while larger pressure loss and fluctuation come together with this effect. The increased pressure loss would generate additional flow resistance, and larger fluctuation of coolant flow would make the mixing effect unstable, as well as would cause flow-induced vibration and grid-to-rod fretting phenomena. The LES method is a viable tool to optimize the spacer grid design and provide transient flow field data for mechanical analysis. Copyright © 2015 by JSME.

关键词： Mixing

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Microstructure and impact toughness of 16MND5 reactor pressure vessel steel manufactured by electrical additive manufacturing

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Journal of Iron and Steel Research International 2020年第8期27卷 992-1004页

作者： Xi-kou He Chang-sheng Xie Li-jun Xiao Ying Luo Di Lu Zheng-dong Liu Xi-tao Wang Institute for Special Steels China Iron and Steel Research InstituteBeijing 100081China Collaborative Innovation Center of Steel Technology University of Science and Technology BeijingBeijing 100083China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610041SichuanChina Southern Additive Science and Technology Co. Ltd.Foshan 528225GuangdongChina

Electrical additive manufacturing can improve manufacturing efficiency and reduce the cost of 16MND5 reactor pres-sure vessel steel. Impact tests were conducted to compare the impact toughness of 16MND5 steels manufactured by the electrical additive manufacturing and conventional forging, respectively. It is found that the impact toughness of electrical additive manufacturing specimen was slightly higher than that of conventional forging specimen. The characterizations of microstructure show that there were large ferrites and carbides in electrical additive manufacturing specimen. The fracture mechanisms of electrical additive manufacturing specimen were that microvoids or microcracks were prone to nucleate at the large ferrite/bainite interface and large carbide/bainitic ferrite interface, where the stress concentration was high. In addi-tion, the block size and high-angle grain boundaries played a vital role in hindering crack propagation of electrical additive manufacturing specimen, helping to improve the impact energy and leading to a low ductile–brittle transition temperature. The results suggest that the electrical additive manufacturing technology was an effective method to enhance the impact toughness of 16MND5 steel.

关键词： Electrical additive manufacturing 16MND5 steel Impact toughness Fracture High-angle grain boundary

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