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

作者： Zheng, Yong Peng, Minjun Ping, A.N. Wei, L.U. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin Heilongjiang China

ISBN: (纸本)9784888983051

The method of characteristics (MOC) with matrix form has more favorable performance compared with its standard application. And the linear algebraic solver affects heavily the computation efficiency. As a Krylov subspace technique, the preconditioned generalized minimal residual algorithm (PGMRES) was proposed to solve the resulting linear system in previous work. To accelerate the iteration process further, the current study utilizes the GPU-based compute unified device architecture(CUDA) technique to program the massively parallel PGMRES. The sparse matrix vector multiplication(SpMV), the most time-consuming operation, was optimized using the coherent visiting and shared memory, which improves the parallel computing performance significantly. Based on the analysis of the numerical feature of the coefficient matrix, two parallel optimization strategies are proposed to deal with the first part of the matrix equation, and the different schemes are utilized to the different parts of the matrix based on the size of row vector, which are expected to improve the throughput of the SpMV operation. The benchmark problem HTTR(i.e. 2D high temperature engineering test reactor benchmark problem) has been simulated to verify the parallel code and evaluate the acceleration performance, and the numerical results demonstrate that the parallel strategies are efficient for the HTTR problem, and the speedup can achieve more than 5.5 times with the optimal strategy. © 2018-2019 by JSME

关键词： Linear systems

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A transient and steady-state network model forannular-wick heat pipes in continuum flow pattern 27

A transient and steady-state network model forannular-wick h...

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

作者： Ma, Yugao Zhong, Ruicheng Zhang, Zhuohua Chai, Xiaoming Huang, Shanfang Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

The heat pipe power system is a potential and near-term space nuclear power system. One of the essential techniques is the high-temperature heat pipe designs with high reliability and heat transfer capacity. In space reactor study, an annular heat pipe, with a liquid annulus between the wick and wall, was proposed to decrease the pressure losses for liquid metal returning and hence, raising the capillary limit of the heat pipes. In this work, a node-based network steady-state and transient model were developed for analyses of these new heat pipes and accuracy improvement in the continuum flow pattern. The model was validated against the experiments and numerical results in the literature. Moreover, the model was applied to calculate the high-temperature heat pipe for space nuclear power system. Preliminary results validate the model for predicting reasonably the thermal-hydraulic features. Copyright © 2018-2019 by JSME

关键词： Heat pipes

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The study of radiation effect on Raman Distributed optical fiber Temperature Sensor

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

作者： Meiqiong Xiang Yanyang Liu Xianguo Qing Zhengxi He Jialiang Zhu Peng He XinLv Science and Technology on Reactor System Design Technology LaboratoryNuclear Power Institute of China Chengdu610213China

The rapid development of big data and artificial intelligence technology is inseparable from the acquisition of massive basic data, and the data is acquired by sensors. The Raman Distributed optical fiber Temperature Sensor (RDTS) is a kind of sensor with many advantages, such as: anti-electromagnetic interference, intrinsic safety, small size, water and high temperature resistance. It has been applied in many fields for fast and continuous temperaturemeasurement. However, due to the rapid attenuation of transmitted light and the temperature error in radiated environment, it has not been applied innuclear field. Therefore, in order to improve the temperature accuracy of RDTS in radiated environment and provide basic data for advanced control. This paper studies the radiation effect on RDTS. Analyzesthe radiation damage mechanism and radiation damage influencing factors of RDTS. Proposes measures to improve the anti-radiation performance and provides four useful calibration methodsfortemperature error. Among them, two calibration methods have been proved right by experiments.

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Evaluation methods of fatigue in nuclear power design

Evaluation methods of fatigue in nuclear power design

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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Wang, Donghui Zeng, Qingna Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China 328# 1st section Chuangshun Dadao Avenue Chengdu610213 China

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A Novel Embedded All-Solid-State Composite Structural Supercapacitor

SSRN

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SSRN 2022年

作者： Zhou, Hanmo Su, Yiyin Zhang, Jing Li, Hao Zhou, Limin Huang, Haitao Department of Applied Physics the Hong Kong Polytechnic University Hung Hom Kowloon Hong Kong Faculty of System Design and Intelligent Manufacturing Southern University of Science and Technology Nanshan District Guangdong Province Shenzhen China Research Center for Humanoid Sensing Zhejiang Lab Kechuang Avenue Zhejiang Province Hangzhou China

Composite structural supercapacitors (CSSs), that provide both mechanical and electrochemical properties, have great potential for applications in electric vehicles, unmanned aerial vehicles and portable electronic devices with reduced overall system weight and volume. In this paper, a novel composite structural supercapacitor device is designed and fabricated, mainly consisting of energy storage units (carbon fiber based flexible devices) and a structural unit (traditional carbon fiber reinforced polymer, CFRP). KOH activated carbon cloth (ACC) and PVA-KOH gel electrolyte are selected as the electrode and electrolyte of the energy storage units, respectively. The specific capacitance, energy density and power density of 1:1 ACC-CSS reach 88 mF×g-1, 9.9 mWh×kg-1 and 445.5 mW×kg-1, respectively. Meanwhile, the flexural strength, flexural modulus and shear strength of 1:1 ACC-CSS achieve 230 MPa, 21 GPa and 8.75 MPa, respectively. In addition, the excellent stability of electrochemical performance of 1:1 ACC-CSS under various external loads is verified by electrochemical three-point bending, electrochemical tensile and electrochemical fatigue tests as well. Experimental results show that even after mechanical failure, 1:1 ACC-CSS can still maintain its electrochemical performance. The combination of good electrochemical and mechanical properties of 1:1 ACC-CSS makes it a promising and competitive candidate for the development of composite structural energy storage devices. © 2022, The Authors. All rights reserved.

关键词： Supercapacitor

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Evaluation and Analysis of RT Suspected Defects on reactor Pressure Boundary Weld

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IOP Conference Series: Earth and Environmental science 2021年第3期632卷

作者： Bo Yang Qiang Deng Cong Chen FengShou Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu China

The control rod driving mechanism (CRDM) is one of the important equipments in the nuclear reactor, and its sealing shell is part of the pressure boundary of the nuclear reactor. The sealing shell is a pressure vessel made of two different material (austenitic stainless steel and nickel base alloy) cylinders welded by dissimilar metal welds. Product production process, in this dissimilar metal weld ray detection found suspected defects show. In this paper, the mechanism and process of the suspected defects are expounded through experimental verification, and it is proved that the suspected defects do not affect the quality of the weld.

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Limit load estimation for finite plates with semi-elliptical defects under tensile loading

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Journal of Physics: Conference Series 2021年第1期1777卷

作者： K K Shi X L Fu H Xie X J Shao J Tian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

The ASME code allows the designer to circumvent many of the problems arising in design by elastic analysis by performing plastic or limit analysis of the component. Unlike elastic analysis, these analyses take account of the redistribution of stress upon yield. In limit analysis, the calculation of the limit load for finite plate with different semi-elliptical defects is an important stage for mechanical safety evaluation. The elastic compensation technique proposed by Mackenzie can be used to provide the limit load of a structure with a finite element method and aims at producing a suitable stress filed for the lower bound theory. In the present study, it is assumed that the finite plate exists three kinds of semi-elliptical cracks with low aspect ratio a/c (a/c=0.1, 0.5 and 1.0), respectively. The improved elastic compensation technique realized by finite element method is applied in the limit load estimation. At last, the results show that the improved elastic compensation technique can be used to obtain the limit value for the finite plate with different semi-elliptical defects under tensile loading. Moreover, the improved elastic compensation technique can be easily implemented in any finite element method and enables the limit load for any complex structure to be obtained swiftly.

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Comparisons of Subgrid-Scale Models for OpenFoam Large- Eddy Simulation

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Journal of Physics: Conference Series 2021年第4期1802卷

作者： Zhipeng Feng Huanhuan Qi Xuan Huang Shuai Liu Jian Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

OpenFOAM is a free, open-source software package that can be used for the solutions of computational fluid dynamics and simulation of various fluid flow processes. Nevertheless, OpenFOAM still lacks default settings and a large number of different numerical schemes and turbulent models should be validated. In this paper, the unsteady flow around a cylinder (Re=3900) is calculated by the large eddy simulation of OpenFOAM. The predictions include the drag and lift coefficient, the pressure distribution around the cylinder, the velocity distribution and Reynolds stress distribution in the wake region, as well as the prediction of the recirculation length and separation angle. Thanks to several simulations, these five subgrid-scale (SGS) models are compared and studied: The Smagorinsky SGS model, wall adaptive local eddy visibility SGS model, dynamic SGS model with Lagrangian averaging, dynamic one equation eddy visibility model, one equation eddy visibility model. The numerical results are verified with the published experimental data.

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Abnormal flow of α particles in heavy-ion collisions at intermediate energies

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Physical Review C 2021年第4期103卷 044607-044607页

作者： G. Qu (曲国峰) Y. Huang (黄宇) D. Peng (彭丹) Z. Xu (徐子虚) W. Lin (林炜平) H. Zheng (郑华) G. Tian (田国玉) R. Han (韩瑞) C. Ma (马春旺) M. Huang (黄美容) P. Ren (任培培) J. Han (韩纪锋 ) Z. Yang (杨振雷) X. Liu (刘星泉) R. Wada Key Laboratory of Radiation Physics and Technology of the Ministry of Education Institute of Nuclear Science and Technology Sichuan University Chengdu 610064China School of Physics Henan Normal University Xinxiang 453007 China School of Physics and Information Technology Shaanxi Normal University Xi'an 710119 China Institute of Modern Physics Chinese Academy of Sciences Lanzhou 730000 China College of Physics and Electronics information Inner Mongolia University for Nationalities Tongliao 028000 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China Cyclotron Institute Texas A&M University College Station Texas 77843 USA

The experimentally observed abnormal α flow behavior from heavy-ion collisions at intermediate energies is investigated using the events of Ca40+Ca40 at 35 MeV/nucleon simulated with an improved antisymmetrized molecular dynamics model with specific consideration of the Fermi motion in the nucleon-nucleon collision process. Its possible origins in the processes from the fragment formation to the experimental flow extraction, i.e., dynamical process, sequential decay, experimental detection, and data analysis, are closely examined. It is found that the observed abnormal α flow behavior originates from the reconstruction of reaction planes in the flow extraction. How the abnormal α flow behavior is generated from the reaction plane reconstruction procedure is discussed.

关键词： Collective flow Low & intermediate energy heavy-ion reactions Models & methods for nuclear reactions

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Effects of heaving condition on characteristics of system under forced circulation and natural circulation 27

Effects of heaving condition on characteristics of system un...

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

作者： Cai, Rong Mi, Zhengpeng Yue, Nina Fang, Hongyu Qiu, Suizheng Wang, Kai Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Chengdu China CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology Chengdu China Xi'an Jiaotong University Xi'an Shaanxi China Department of Nuclear Engineering and Management School of Engineering University of Tokyo Tokyo Japan

ISBN: (纸本)9784888983051

Under ocean conditions, the nuclear power plants have complex movements due to the effects of wind and wave. The thermal-hydraulic characteristics of the nuclear power plants under ocean conditions are influenced by its various motions. To ensure the safety of the reactor, the research on the effects of various motions on its thermal-hydraulic characteristics are necessary. The previous research has shown that heaving condition has more serious effects on the thermal-hydraulic characteristics of the nuclear power plants. In this paper, a thermal-hydraulic system code for ocean conditions named STAC was developed, and the STAC code was validated by test data. The effects of heaving conditions on the thermal-hydraulic characteristics of the system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under heaving condition fluctuate periodically, and the fluctuation period is the same as the heaving period. The delay of the variation of circuit flow and the thermal inertia of the fuel material cause a phase difference between the fluctuations of the thermal parameters of reactor core. Heaving motion has less effect on the nuclear power system under forced circulation than that under natural circulation. As the amplitude of the acceleration increases, the fluctuation amplitudes of thermal-hydraulic parameters in the nuclear power system increase. Under heaving condition with short period, the fluctuation amplitudes of thermal-hydraulic parameters are affected by the heaving period. With the increase of the fluctuation period, the fluctuation amplitudes of thermal-hydraulic parameters increase gradually, and finally approach certain values. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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