检索结果-内蒙古大学图书馆

您好，读者！请登录

内蒙古大学图书馆

首页
概况
党建
资源
服务
科研支持
- 论文收录引用证明
- 科技查新
知识产权
档案馆
帮助

咨询与建议

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

您的常用邮箱：*

您的手机号码：*

问题描述：

当前已输入0个字，您还可以输入200个字

全部搜索
期刊论文
图书
学位论文
标准
纸本馆藏
外文资源发现
数据库导航
超星发现

高级检索

时间限定

出版年份：

文献类型

图书期刊文献学位论文多媒体

馆藏选择

电子馆藏纸本馆藏

核心期刊

全部期刊 SCI 收录期刊 SSCI 收录期刊 EI 收录期刊 CSCD 收录期刊 CSSCI 收录期刊

语言

中文英文

文献类型

期刊文献图书学位论文标准纸本馆藏

帮助

文字说明：

T=题名（书名、题名），A=作者（责任者），K=主题词，P=出版物名称，PU=出版社名称，O=机构（作者单位、学位授予单位、专利申请人），L=中图分类号，C=学科分类号，U=全部字段，Y=年（出版发行年、学位年度、标准发布年）

检索规则说明：

AND代表“并且”；OR代表“或者”；NOT代表“不包含”；(注意必须大写,运算符两边需空一格)

检索范例：

范例一：(K=图书馆学 OR K=情报学) AND A=范并思 AND Y=1982-2016
范例二：P=计算机应用与软件 AND (U=C++ OR U=Basic) NOT K=Visual AND Y=2011-2016

分类表

所选分类

>> <<

限定检索结果

文献类型

690 篇 会议
585 篇 期刊文献
1 册 图书

馆藏范围

1,276 篇 电子文献
0 种 纸本馆藏

日期分布

学科分类号

1,024 篇 工学
- 358 篇 动力工程及工程热...
- 279 篇 计算机科学与技术...
- 267 篇 核科学与技术
- 247 篇 软件工程
- 228 篇 力学（可授工学、理...
- 219 篇 化学工程与技术
- 186 篇 机械工程
- 162 篇 材料科学与工程（可...
- 147 篇 安全科学与工程
- 121 篇 控制科学与工程
- 117 篇 冶金工程
- 112 篇 电气工程
- 78 篇 土木工程
- 70 篇 电子科学与技术（可...
- 64 篇 仪器科学与技术
- 62 篇 交通运输工程
- 49 篇 信息与通信工程
- 44 篇 航空宇航科学与技...
- 41 篇 生物工程
- 36 篇 光学工程
710 篇 理学
- 372 篇 物理学
- 239 篇 化学
- 238 篇 数学
- 73 篇 统计学（可授理学、...
- 55 篇 生物学
- 27 篇 海洋科学
- 24 篇 系统科学
130 篇 管理学
- 116 篇 管理科学与工程(可...
23 篇 医学
14 篇 经济学
10 篇 法学
9 篇 农学
6 篇 教育学
2 篇 军事学
2 篇 艺术学
1 篇 文学

主题

35 篇 nuclear power pl...
16 篇 finite element m...
11 篇 pressurized wate...
10 篇 heat transfer
9 篇 two phase flow
9 篇 heat flux
9 篇 optimization
9 篇 predictive model...
9 篇 silicon carbide
8 篇 simulation
8 篇 sensitivity anal...
8 篇 analytical model...
8 篇 training
7 篇 deep learning
7 篇 computational mo...
7 篇 control systems
7 篇 loss of coolant ...
7 篇 sensors
6 篇 corrosion resist...
6 篇 computational fl...

机构

409 篇 science and tech...
57 篇 science and tech...
39 篇 department of en...
38 篇 school of system...
32 篇 nuclear power in...
26 篇 tianjin key labo...
26 篇 cas key laborato...
25 篇 nuclear power in...
16 篇 national demonst...
12 篇 henan key labora...
11 篇 key laboratory o...
11 篇 fundamental scie...
11 篇 shenzhen key lab...
11 篇 school of nuclea...
10 篇 school of mechat...
10 篇 peng cheng labor...
9 篇 national key lab...
9 篇 college of nucle...
9 篇 key laboratory o...
8 篇 national key lab...

作者

39 篇 deng jian
23 篇 wang kan
23 篇 li qing
21 篇 yu hongxing
17 篇 qiu suizheng
17 篇 tian wenxi
15 篇 peng xingjie
14 篇 su guanghui
13 篇 chen zhi
13 篇 chen ping
13 篇 yao dong
12 篇 huang shanfang
12 篇 zhang dalin
12 篇 gong helin
12 篇 li jishun
11 篇 yu yingrui
11 篇 zhang xu
11 篇 chai xiaoming
11 篇 ma yugao
10 篇 wu dan

语言

1,190 篇 英文
46 篇 中文
31 篇 其他
3 篇 日文
1 篇 法文

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

共 1276 条记录，以下是131-140 订阅

全选清除本页清除全部题录导出标记到"检索档案"

详细简洁

排序：

A combined reactivity control pattern on a floating nuclear power plant 27

A combined reactivity control pattern on a floating nuclear ...

引用

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

作者： Zhou, Bingyan Song, Danrong Chen, Zhang Chai, Xiaoming Xiang, Hongzhi Hu, Yuying Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

In this study, a new combined reactivity control pattern is presented to lower the power peaking factor on a floating nuclear power plant. In the pattern the whole cycle length is divided into several boron-adjusting periods. During one boron-adjusting period, the soluble boron concentration remains unchanged and the core reactivity change is controlled by control rods. Boron concentration would only be changed at the end of a boron-adjusting period. In the whole cycle the boron concentration would be changed for two to four times, which is acceptable for a floating nuclear power plant. To get an optimized solution, a model of the combined reactivity control pattern was established, which contained boron concentration and control rod pattern (CRP) as coupled variables. A particle swarm optimization with novel mutation operator (PSONM) was applied to optimize the solution. The model was practiced based on a small modular reactor ACP100 core. The solution was optimized separately for equal division of two, three, four and five boron-adjusting periods. The results show that the combined reactivity control pattern can get a lower power peak factor than the original rod-controlled-only pattern without shortening the cycle length. Therefore, the safety of the floating nuclear power plant can be improved. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

来源：评论

学校读者我要写书评

暂无评论

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

来源：评论

学校读者我要写书评

暂无评论

Study on Acoustic-Induced Vibration (AIV) by reactor coolant pump at secondary support structure 48

Study on Acoustic-Induced Vibration (AIV) by reactor coolant...

引用

48th International Congress and Exhibition on Noise Control Engineering, INTER-NOISE 2019 MADRID

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

ISBN: (纸本)9788487985317

The reactor coolant pump is the core equipment of nuclear power stations. During the operation of reactor coolant pump, acoustic pulsations are generated due to nonuniform flow within the reactor coolant pump and pass through the main pipe and coolant to the structure in the reactor. The acoustic pulsations can cause the vibration of the structure in the reactor, which is regarded as one of the main reasons of fatigue failure of the structure. In this paper, the finite element model of secondary support structure will be built and verified by test, then the pressure pulsations relation between the reactor coolant pump outlet and the secondary structure will be investigated, the pressure pulsation caused by the reactor coolant was calculated. Finally, the acoustic-induced vibration response of secondary support structure caused by the reactor coolant pump were obtained by harmonic analysis and high circle fatigue analysis will be done on the basis of the results. This study is helpful to understand the acoustic-induced vibration of the reactor coolant pump and to improve the hydraulic design of the reactor coolant pump. © INTER-NOISE 2019 MADRID - 48th International Congress and Exhibition on Noise Control Engineering. All Rights Reserved.

关键词： Harmonic analysis

来源：评论

学校读者我要写书评

暂无评论

Stress intensity factors for penetrate radial cracks in reactor coolant pump flywheel by weight function method 14

Stress intensity factors for penetrate radial cracks in reac...

引用

14th International Conference on Fracture, ICF 2017

作者： Shi, K.K. Zheng, B. Gao, S.Q. Chen, J.G. Lv, X. Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Stress intensity factors for penetrate radial crack at center of keyway and at keyway corner under the plane stress, the plane strain and the three dimensional, based on weight function method, is determined by finite element analysis. Two load cases of pin loading and uniform loading are considered in the present study. About stress intensity factors, finally, an approximate solution and results from weight function method are compared. Several loading are considered that lead to exact solutions which agree with the literature. The accuracy of the analysis has been examined using the finite element method results and were compared to existing solutions for uniform loading in the literature for special geometries, indicating an excellent agreement. © 2017 ICF 2017 - 14th International Conference on Fracture. All rights reserved.

关键词： Stress intensity factors

来源：评论

学校读者我要写书评

暂无评论

Application of optimus in sensitivity analysis of dynamic response of reactor coolant system 27

Application of optimus in sensitivity analysis of dynamic re...

引用

27th International Congress on Sound and Vibration, ICSV 2021

作者： Yanli, Yuan Xianhui, Ye WenZheng, Zhang Lijuan, Li Zhenyu, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9788378807995

In this paper, the sensitivity of the dynamic response of reactor coolant system to the input parameters is investigated through OPTIMUS integrated platform, taking the stiffness of the dampers which support the steam generator and the main pump in the reactor coolant system as the input variables. OPTIMUS is a kind of software developed by Noesis Solutions Company, by which the multidisciplinary analysis process can be integrated for collaborative optimization design. OPTIMUS allows users to graphically establish workflows in which they define input files, output files, input parameters, output parameters and analysis tools that are needed to be called and the logical relationships between controls in the process. The dynamic response analysis of reactor system is the key process in the mechanical evaluation. With the continuous improvement of seismic design benchmark and reactor design requirements, the problem of structural stress over-limit is becoming more and more prominent, and the repeated iteration of design to evaluation becomes an inevitable trend. It is necessary to examine the sensitivity of the dynamic response to the key design parameters in the reactor system and find out the design parameters which have a great impact on the reactor dynamic response to shorten the design cycle and improve the design efficiency. In this paper, the sensitivity of the output parameters to the input parameters are investigated by OPTIMUS, considering the stiffness of dampers supporting the steam generator and the main pump as the input parameters, taking the supporting loads of the main equipments as the output parameters. Not only can ANSYS simulation calculations be automatically carried out on the OPTIMUS, but also can the output data be disposed automatically and rapidly. The sensitivity analysis results can be used for the determination of design variables in the subsequent optimization analysis. © "Advances in Acoustics, Noise and Vibration - 2021" Proceedings of

关键词： Sensitivity analysis

来源：评论

学校读者我要写书评

暂无评论

Development of three-dimensional neutron kinetics code based on high order nodal expansion method in hexagonal-Z geometry 26

引用

2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Chao, Guo Yu, Liu Hangxing, He Luguo, Liu Xiaoyu, Wang Sufang, Xin Peiying, Li Xiaoli, Wu Hongsheng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780791851456

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry(HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated. Copyright © 2018 ASME.

关键词： Runge Kutta methods

来源：评论

学校读者我要写书评

暂无评论

Multiple branches coupling property and design of PRHRs

Multiple branches coupling property and design of PRHRs

引用

2018 Transactions of the American Nuclear Society, ANS 2018 and Embedded Topical Meeting Nuclear Fuels and Structural Materials

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

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. Copyright © 2018 AMERICAN NUCLEAR SOCIETY, INCORPORATED, LA GRANGE PARK , ILLINOIS 60526Printed in USA.

关键词： Heat transfer

来源：评论

学校读者我要写书评

暂无评论

The preliminary performance study of ATF with M3 fuel and Zr-4 cladding 24

引用

2016 24th International Conference on Nuclear Engineering, ICONE 2016

作者： Chunyu, Yin Teng, Tu Tang, Changbing Jiao, Yongjun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China ChengduSichuan China

ISBN: (纸本)9780791850015

Metal matrix microencapsulated (M3) fuel is one of the research directions on Accident Tolerant Fuel (ATF). In this article, it provides one of ATF design which consists of BISO (Bistructural ISOtropic) particles embedded in a zirconium alloy matrix, and the cladding coating with silicon carbon (SiC). The temperature distribution of the ATF element has been built, and then the center temperature has also been calculated based on the operation parameters of the large-scale pressurized-water reactor. Simultaneity, the several factors of fuel failure is preliminary analyzed and calculated, especially the pressure shell failure mechanism. Copyright © 2016 by ASME.

关键词： Metallic matrix composites

来源：评论

学校读者我要写书评

暂无评论

Validation of sop procedure for dealing with FWLB 22

引用

2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Yaou, Shen Wei, Chen Daishun, Huang Guijun, Leng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9780791845936

The State Oriented Procedure (SOP) is developed to improve the existing emergency operation procedures, namely, the Event Oriented Procedure (EOP). It is developed based on the finite Nuclear Steam Supply system (NSSS) states instead of an infinite number of possible events, therefore it can cover all of the events/accidents including the accidents caused by more than one initial event. On the other hand, the operator can take actions without diagnosing the events/accidents according to SOP, and thus the human mistakes which will possible happen during the events/accidents diagnosing can be avoided. However, as a generalized procedure system, the SOP needs a lot of validations based on various events/accidents. In this work, an analysis of a postulated Feed Water Line Break (FWLB) accident for LingAo phase II NPP based on the SOP is given and both the automatic protections and operator manual actions required by the SOP are modeled with an advanced, two-fluid, thermal-hydraulic code called CATHARE. The initial conditions and assumptions are chosen to get a severe state of the NPP. Since the EOP addresses the given accident perfectly, the results are then compared with those based on the EOP for a similar NPP to validate the SOP. The main parameters relevant with FWLB of this NPP are the same with LingAo phase II. Therefore, there is no need to perform analyses based on the EOP for LingAo phase II. The results demonstrate that the main operator manual actions required by the SOP are similar to those required by the EOP, and the operator can take the plant to a controllable safe state according to the SOP procedures. Therefore, the SOP is effective and reasonable for FWLB accident. Copyright © 2014 by ASME.

关键词： Nuclear engineering

来源：评论

学校读者我要写书评

暂无评论

Anti-interference Study of Wireless Sensor Networks in Nuclear Power Plants 1

引用

7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Deng, Zhi-guang Wu, Qian Lv, Xin Xiang, Mei-qiong Xu, Tao Qin, Yue Sun, Yong-sheng Peng, Ye-shun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

The compatibility and anti-interference capability of wireless sensor networks within nuclear power plants in complex electromagnetic environments are the most essential elements to ensure reliable operation and effective combat of equipment within nuclear power plants in harsh environments. Interference signals generated by electrical, rotating and communication equipment and sensors attached to nuclear power plants, as well as malicious enemy interference sources, cause their electromagnetic environment to become increasingly complex. For the complex interference environment of nuclear power plant compartment, this paper firstly adopts supervised and unsupervised learning for interference cognition, and after obtaining the features of wireless interference sequences, anti-interference decision based on deep reinforcement learning and array antenna multi-domain anti-interference based on intelligent decision are implemented. From the intelligent interference simulation based on the physical transmission system, it can be seen that the proposed method can effectively identify and resist typical interference signals, which can bring reference to the anti-interference of nuclear power plants. © 2023, 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

没有更多数据了...

全选清除本页清除全部题录导出标记到“检索档案”

共128页 << < 10 11 12 13 14 15 16 17 18 19 > >>

检索报告对象比较合并检索0

隐藏清空

合并搜索

回到顶部

执行限定条件

内容：

评分：

请选择保存的检索档案：

请选择收藏分类：

订阅名称：

通借通还

温馨提示：

图书名称：

借书校区：

取书校区：

手机号码：

邮箱地址：

一卡通帐号：

电话和邮箱必须正确填写，我们会与您联系确认。

联系人：

所在院系：

联系邮箱：

联系电话：

内蒙古自治区呼和浩特市赛罕区大学西街235号邮编: 010021

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：