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

作者： Wu, Dan Deng, Jian Du, Sijia Qian, Libo 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 an over pressure accident, one or more pressurizer safety(or relief) valves will open due to the rapid pressure rise process. Once the safety(or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety(or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety(or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety(or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety(or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety(or relief) valve setpoint is very important to the discharge force analysis results. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Liquids

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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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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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Research on Output Torque Optimization of Ultrasonic Motor Based on Genetic Algorithm 17th

Research on Output Torque Optimization of Ultrasonic Motor B...

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17th Annual Conference of China Electrotechnical Society, CES 2022

作者： Li, Xiaoyang Tang, Jiankai Yang, Bo Wu, Hao Xu, Qiwei Yang, Yun The State Key Laboratory of Power Transmission Equipment & System Security and New Technology Chongqing University Chongqing400030 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9789819904501

Aiming at the problem of small output torque in the design process of traveling to wave ultrasonic motor, this paper proposes a method of optimizing the output torque of traveling to wave ultrasonic motor based on multi-objective genetic algorithm. It can effectively improve the output torque of the motor. Firstly, the structure and operation mechanism of the motor is introduced, and the key factors affecting the output torque of the motor is analyzed reasonably;then the parameterized model of the stator of the ultrasonic motor is established by WORKBENCH, the sample points are randomly selected in the design variable space, and the combination of each sample point is obtained. The modal analysis and harmonic response analysis of the stator structure was carried out to establish the response surfaces model of the stator;finally, the global optimal solution was obtained based on the response surface model by using the multi-objective genetic algorithm (MOGA). The results show that the output torque of the ultra-sonic motor can be increased by 8% before and after optimization. © 2023, Beijing Paike Culture Commu. Co., Ltd.

关键词： Stators

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Wear Characteristics of the Nuclear Control Rod Drive Mechanism(CRDM)Movable Latch Serviced in High Temperature Water

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Chinese Journal of Mechanical Engineering 2022年第2期35卷 111-120页

作者： Tianda Yu Guozhong Fu Yanqing Yu Liting Zhu Maofu Liu Wei Li Qiang Deng Zhenbing Cai The State Key Laboratory of Power Transmission Equipment&System Security and New Technology Chongqing UniversityChongqing 4000442China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610041China Tribology Research Institute Southwest Jiaotong UniversityChengdu 610031China Sichuan Huadu Nuclear Equipment Manufacturing Co. LtdDujiangyan 611830China

The current research of nuclear control rod drive mechanism(CRDM)movable latch only makes a simple measurement of wear *** wear volume and difference in various claw surfaces are ignored and the degradation mechanism of each claw surface is not *** this paper,a detailed degradation analysis was carried out on each claw surface of movable latch combined with wear result and worn *** indicate that the boundary of carbide is preferred for corrosion because carbide presents a nobler Volta potential compared to the metal matrix or boundary *** to the oscillation of drive shaft between the claw surfaces of movable latch,the dominant wear mechanism on the upper surface of claw(USC)and lower surface of claw(LSC)is plastic deformation caused by impact *** impact wear will cause the fragmentation of carbides because of the high hardness and low ductility of *** promotes the broken carbides to fall off from the metal *** generated fine carbides(abrasive particles)cause extra abrasive wear on USC when the movable brings the drive shaft upward or *** a result,USC has a higher wear volume than *** research proposes a method to evaluate the wear on the whole movable latches using a 3D full-size scanner.

关键词： CRDM Movable latch Degradation analysis Wear characteristics Carbide

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Generative API usage code recommendation with parameter concretization

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science China(Information sciences) 2019年第9期62卷 51-72页

作者： Chi CHEN Xin PENG Jun SUN Zhenchang XING Xin WANG Yifan ZHAO Hairui ZHANG Wenyun ZHAO School of Computer Science Fudan University Shanghai Key Laboratory of Data Science Fudan University Shanghai Institute of Intelligent Electronics & Systems Pillar of Information System Technology and Design Singapore University of Technology and Design Research School of Computer Science Australian National University

Many programming languages and development frameworks have extensive libraries(e.g., JDK and Android libraries) that ease the task of software engineering if used effectively. With numerous library classes and sometimes intricate API(application programming interface) usage constraints, programmers often have difficulty remembering the library APIs and/or using them correctly. This study addresses this problem by developing an engine called DeepAPIRec, which automatically recommends the API usage *** to the existing proposals, our approach distinguishes itself in two ways. First, it is based on a tree-based long short-term memory(LSTM) neural network inspired by recent developments in the machinelearning community. A tree-based LSTM neural network allows us to model and reason about variable-length,preceding and succeeding code contexts, and to make precise predictions. Second, we apply data-flow analysis to generate concrete parameters for the API usage code, which not only allows us to generate complete code recommendations but also improves the accuracy of the learning results according to the tree-based LSTM neural network. Our approach has been implemented for supporting Java programs. Our experimental studies on the JDK library show that at statement-level recommendations, DeepAPIRec can achieve a top-1 accuracy of about 37% and a top-5 accuracy of about 64%, which are significantly better than the existing approaches. Our user study further confirms that DeepAPIRec can help developers to complete a segment of code faster and more accurately as compared to IntelliJ IDEA.

关键词： code recommendation API deep learning data dependency parameter concretization

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