检索结果-内蒙古大学图书馆

Study on flow modes of open natural circulation with low height difference under low power conditions

International Journal of Advanced Nuclear reactor design and technology 2020年 2卷 10-14页

作者： Yi, Li Jianchuang, Sun Hang, Peng Cheng, Xiang Quan, Biao Cao, Xiaxin Jian, Zhou Ming, Ding Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin150001 China

Open natural circulation systems are widely applied in the field of energy and chemical industries. Under some special conditions, open natural circulation systems with low height difference need to be applied, but research on this type of systems is not sufficient. In this paper, the flow modes of sucha natural circulation system were experimentally studied at low power level by adopting steam heating. In addition, the flow modes and physical phenomena of the natural circulation system were analyzed in detail under different heating power conditions. It is found that there are four flow modes at different power levels in the natural circulation system due to the influence of non-condensable gas and subcooled boiling. The inlet and outlet temperatures of heating section are analyzed in different flow modes. The result shows that the outlet temperature of heating section can be used as a judgment criterion for different flow modes. © 2019

关键词： Chemical industry

来源：评论

学校读者我要写书评

暂无评论

PREDICTION OF CRITICAL HEAT FLUX BASED ON RANDOM FOREST ALGORITHM 29

PREDICTION OF CRITICAL HEAT FLUX BASED ON RANDOM FOREST ALGO...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Sun, Zhuo Han, Mingyang Yuan, Peng Gong, Lei Chen, Di Yuan, Guanghui Zhang, Shuang Nuclear Power Institute of China Chengdu China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886410

The accurate prediction of critical heat flux(CHF) is of great significance to the safe operation of the reactor. At present, most of the traditional methods for predicting CHF are based on relational fitting in mathematical form, which are not accurate enough and may can not taking full advantage of experimental data. Therefore, a prediction model based on random forest algorithm (RF) is proposed, the model is based on CHF 2006 look-up table, and the accuracy and smoothness of the model have been verified when predicting CHF. In this paper, the variation trend of CHF predicted value and actual value with pressure, mass flow rate and thermodynamic quality is given. It has been verified that the error of this method is acceptable, the predicted value is in good agreement with the experimental value, and conforms to the corresponding physical laws, which can provide a reference for the development of high-precision CHF prediction models or CHF-related experiments. Copyright © 2022 by ASME.

关键词： Heat flux

来源：评论

学校读者我要写书评

暂无评论

Research on the application of Cobalt-Based Alloy Laser Additive Manufacturing in reactor Internals 29

Research on the application of Cobalt-Based Alloy Laser Addi...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Wang, Qingtian Cao, Rui Zhang, Xianjun Science and Technology on Reactor System Design Technology Laboratory Chengdu China Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886458

In view of the quality problems of manual TIG coating of cobalt-based alloy, the Laser Additive Manufacturing process was studied, especially the optimization of process parameters. Then a series of experiments were carried out on the cobalt-based alloy Additive Manufacturing layer and compared with the traditional manual TIG coating layer. The results show that the microstructure, the hardness, the wear resistance and corrosion resistance are better than. Copyright © 2022 by ASME.

关键词： Microstructure

来源：评论

学校读者我要写书评

暂无评论

Study on melting behavior and related influential factors of shroud and basket of acp1000 during a severe accident

Study on melting behavior and related influential factors of...

引用

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

作者： Hou, Li-Qiang Zhu, Da-Huan Wu, Qing Deng, Jian Wu, Xiao-Li Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu of Sichuan Prov 610213 China

ISBN: (纸本)9780894487552

The focusing effect of the metal molten pool plays an important role in elevating the validity of the In-Vessel Retention (IVR) during a severe accident while the melting of the shroud and basket can contribute to the formation of the metal molten pool. Therefore, the study on the melting behavior of the shroud and basket can offer technical support for the validity analysis of IVR. The method of computational fluid dynamics (CFD) has been used to establish a two-dimensional calculation model of the 1/8 core of ACP1000 to study the melting behavior of the shroud and basket during the severe accident of large LOCA. The core has been divided into ten nodes in the axial direction and the radiation model and the solidification and melting model have been considered during the calculation. The results show that the shroud at six out of ten nodes and the basket at five out of ten nodes have melted totally before the first significant migration of the core and that the basket always starts to melt after the shroud has melted totally at the same node. © 2018 American Nuclear Society. All rights reserved.

关键词： Melting

来源：评论

学校读者我要写书评

暂无评论

design Method of Human-Computer Interactive Interface of MTS Based on DCS of Nuclear Power Plant 5th

Design Method of Human-Computer Interactive Interface of MTS...

引用

5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Dang, Li-Jun Huang, Jun Ye, Qi Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan610213 China

ISBN: (纸本)9789811634550

MTS (maintenance and test software), as an important human-computer interactive tool in DCS of nuclear power plant, plays a crucial role in configuration and management of the engineering applied data, engineering design, system debugging and operational maintenance. With the advent of information age, industrial designers pay more and more attention to interactive design, and user attaches increasing importance to the human-machine interactive experience of products, which puts forward new requirements for the human-machine interactive interface of MTS. This paper takes the human-computer interaction visual design of MTS as the core, through analyzing the human-computer interaction model and requirements of MTS, formulates the overall design scheme of the human-computer interaction system of MTS. According to the overall scheme, a set of human-computer interaction visual design method applied to MTS of DCS in nuclear power plant is proposed. The method and criterion of human-computer interaction design proposed in this paper is very important for the design human-computer interaction of nuclear power plant application software, and also points out the direction of human-computer interaction optimization of MTS mentioned in this paper. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

来源：评论

学校读者我要写书评

暂无评论

Semi-empirical and semi-quantitative lightweight shielding design method

引用

Nuclear science and Techniques 2023年第3期34卷 109-124页

作者： Song-Chuan Zheng Qing-Quan Pan Huan-Wen Lv Song-Qian Tang Xiao-Jing Liu College of Smart Energy Shanghai Jiao Tong UniversityShanghai 200240China School of Nuclear Science and Engineering Shanghai Jiao Tong UniversityShanghai 200240China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610200China

The lightweight shielding design of small reactors is a popular research *** on a small helium-xenon-cooled solid reactor,the effects of neutron and photon shielding sequence and the number of shielding layers on the radiation dose were first *** was found that when photons were shielded first and the number of shielding layers was odd,the radiation dose could be significantly *** reduce the weight of the shielding body,the relative thickness of the shielding layers was optimized using the genetic *** optimized scheme can reduce the radiation dose by up to 57%and reduce the weight by 11.84%.To determine the total thickness of the shielding layers and avoid the local optimal solution of the genetic algorithm,a series of formulas that describes the relationship between the total thickness and the radiation dose was developed through large-scale calculations.A semi-empirical and semi-quantitative lightweight shielding design method is proposed to integrate the above shielding optimization method that verified by the Monte Carlo ***,a code,SDIC1.0,was developed to achieve the optimized lightweight shielding design for small *** was verified that the difference between the SDIC1.0 and the RMC code is approximately 10%and that the computation time is shortened by 6.3 times.

关键词： Small reactor Lightweight Shielding calculation Genetic algorithm

来源：评论

学校读者我要写书评

暂无评论

Safety assessment of UN/Zr fuel with SiC cladding under large break loss-of-coolant accident 14

Safety assessment of UN/Zr fuel with SiC cladding under larg...

引用

14th International Nuclear Fuel Cycle Conference, GLOBAL 2019 and Light Water reactor Fuel Performance Conference, TOP FUEL 2019

作者： Qian, Libo Yu, Hongxing Sun, Yufa Deng, Jian Wu, Dan Ding, Shuhua Chen, Wei Li, Zhongchun Chen, Xi Liu, Lili Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan610213 China

Accident Tolerant Fuel (ATF), which is aiming at upgrading safety characteristics of nuclear fuel and cladding under design basis and design extension conditions, has drawn more and more attention since Fukushima accident. By virtue of improved basic properties and oxidation resistance comparing with conventional UO2-Zr, UN(Particle)/Zr(Matrix) fuel pin and SiC cladding (UN/Zr-SiC) is expected to slow down accident progression and to provide wider margin of coping time. In present paper, the improved performance of UN/Zr-SiC fuel system on the Large Break Loss-Of-Coolant-Accident (LB LOCA) is analyzed by Appendix K version of ARSAC code (ARSAC-K). For this reason, the corresponding models in ARSAC-K are modified to achieve the goal: 1) thermal physical properties of SiC;2) equivalent thermal physical properties of UN/Zr composite fuel;3) passive oxidation model of SiC under pure steam condition based on modified Deal-Grove model together with linear volatilization model of SiO2 based on heat/mass transfer analogy. Obtained quantitative results show promising safety enhancement in terms of peak cladding temperature (PCT): the PCT of UN/Zr-SiC fuel is much lower than that of UO2-Zr fuel. Then due to the great safety margin of UN/Zr-SiC fuel under LB LOCA condition, an exploratory investigation of the reduction of safety injection capacity and the increase of reactor nominal power under LB LOCA condition is presented. The results show that from the LB LOCA point of view, UN/Zr-SiC fuel has the potential can to be utilized to reduce safety injection capacity or increase reactor nominal power without reducing the safety margin. Therefore, UN/Zr-SiC fuel is demonstrated to have positive effects on safety and capability of reactor design optimization. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water reactor Fuel Performance *** rights reserved.

关键词： Silicon carbide

来源：评论

学校读者我要写书评

暂无评论

INVESTIGATION ON THE SIMULATION OF TURBULENT HEAT TRANSFER OF LIQUID METAL FLOW IN A NARROW RECTANGULAR CHANNEL 29

INVESTIGATION ON THE SIMULATION OF TURBULENT HEAT TRANSFER O...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zhang, Xue Yu, Hongxing Deng, Jian Du, Sijia Zhang, Xilin Zou, Ziqiang Feng, Wenpei Zhang, Yaoxiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Hefei China

ISBN: (纸本)9780791886434

Liquid metal has the advantages of high boiling point and strong thermal conductivity and is selected as the coolant for a new generation nuclear system. In addition, the narrow rectangular channel is used in the small reactors or research reactors due to its compact structure and high heat transfer capacity. In order to accurately simulate the turbulent heat transfer of liquid metals in narrow rectangular channels, based on the relevant experimental data, this paper explored suitable numerical simulation methods. First of all, according to the research object, the appropriate experimental data was selected, and the geometric modeling, mesh division, and the mesh independence verification were carried out. It was ensured that the y+ meets the requirements of each turbulence model and wall treatment model. Several typical turbulence models and the wall treatment models were evaluated. For the low Prandtl number fluid such as liquid metal, it is necessary to select a suitable turbulent Prandtl number model in the RANS simulation. The sensitivity of the calculation results to the turbulent Prandtl number was analyzed. Finally, the existing turbulent Prandtl number models were evaluated. The suitable at different Pe was given based on experimental value, so that the turbulent heat transfer of liquid metal in a narrow rectangular channel can be accurately simulated. Copyright © 2022 by ASME.

关键词： Liquid metals

来源：评论

学校读者我要写书评

暂无评论

Simulation analysis on the pressurizer degassing sub-system of module small reactor (ACP100)

Simulation analysis on the pressurizer degassing sub-system ...

引用

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

作者： Zhiyun, Cai Yun, Ren Jianyong, Lai Yulong, Zhang Hang, Liu Xianghong, Liu Minghao, Liu Zhu, Ye Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu610041 China

ISBN: (纸本)9780791883778

A pressurizer degassing sub-system is designed in Module Small reactor(ACP100), which can be used to vent the noncondensable gas dissolved in reactor Coolant during plant shutdown period by opening the remote degassing valve installed on top of the pressurizer. In order to verify the transient characteristic during the whole degassing process, a dedicated transient degassing program is developed based on the improved non-equilibrium multi-region model and the transient degassing model, which is used then to carry out a simulation analysis on this process. Also, the boundary condition models including pressurizer heater power model, pressure spray model, surge line flow model and critical flow model for degassing valve are established. The simulation results show that, under ultimate condition of hot-zero power operation, during the whole degassing process, the Pressurizer's pressure drops to a maximum 0.038MPa, and the water level is raised 0.0637m above the normal level, which means, both of the pressure and water level are within the normal operation band and shall not initiate any safety signal. Meanwhile, the whole transient process lasts about 16 minutes, and then it enters into the stable degassing period. The gas concentration vs time curve is very smooth. It takes about 5.8 hours to remove the gas dissolved in reactor Coolant from 35 ml(STP)/kg to 15 ml(STP)/kg. The analysis shows the Pressurizer Degassing Sub-system is safety, effectively and reliable. Copyright © 2020 ASME

关键词： Plant shutdowns

来源：评论

学校读者我要写书评

暂无评论

Optimization design for the configuration parameters of helical-coil steam generator volume

Optimization design for the configuration parameters of heli...

引用

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

作者： Fang, Hongyu Wang, Li Zhu, Dahuan Cai, Rong Jiang, Xiaowei Zhang, Dan Qian, Libo Wu, Dan Science and Technology on Reactor System Design Technology Laboratory Chengdu China Sun YAT-SEN University Guangzhou China

ISBN: (纸本)9780791883785

With the development of the nuclear energy industry, small modular reactors (SMRs) have become an important option in China's energy development due to their advantages in terms of safety and economics. The helical-coil steam generator is an important part of SMR, and the structure parameters need to be optimized urgently to reduce costs and improve safety. In this paper, various thermohydraulic parameters were used as indicators for the design quality of the helical-coil steam generator, such as the volume of the equipment, velocity of the flow, maximum temperature of the tubes, and natural circulation capability. The optimization of these target parameters has important theoretical and practical values. The genetic algorithm method is widely used to processes data efficiently, and it is able to reach a optimum design rapidly, which makes it a good choice for the optimization of parameters. The results show that this method could efficiently reduce the volume of the helical coil steam generator. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Genetic algorithms

来源：评论

学校读者我要写书评

暂无评论

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案：

请选择收藏分类：

通借通还

建议与咨询 留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

时间限定

文献类型

馆藏选择

核心期刊

语言

文献类型

帮助

文字说明：

检索规则说明：

检索范例：

分类表

所选分类

限定检索结果

文献类型

馆藏范围

日期分布

学科分类号

主题

机构

作者

语言

请选择保存的检索档案： 新增检索档案 确定 取消

请选择收藏分类： 新增自定义分类 确定 取消

通借通还

建议与咨询留下您的常用邮箱和电话号码，以便我们向您反馈解决方案和替代方法

请选择保存的检索档案：

请选择收藏分类：