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

1,073 篇 会议
906 篇 期刊文献

馆藏范围

1,979 篇 电子文献
0 种 纸本馆藏

日期分布

学科分类号

1,479 篇 工学
- 433 篇 计算机科学与技术...
- 395 篇 动力工程及工程热...
- 368 篇 软件工程
- 301 篇 力学（可授工学、理...
- 270 篇 机械工程
- 266 篇 化学工程与技术
- 255 篇 核科学与技术
- 236 篇 电气工程
- 223 篇 材料科学与工程（可...
- 198 篇 控制科学与工程
- 197 篇 电子科学与技术（可...
- 151 篇 安全科学与工程
- 134 篇 冶金工程
- 116 篇 信息与通信工程
- 115 篇 航空宇航科学与技...
- 104 篇 仪器科学与技术
- 103 篇 土木工程
- 71 篇 交通运输工程
- 63 篇 生物工程
- 51 篇 光学工程
945 篇 理学
- 530 篇 物理学
- 339 篇 数学
- 278 篇 化学
- 85 篇 统计学（可授理学、...
- 80 篇 生物学
- 45 篇 系统科学
183 篇 管理学
- 156 篇 管理科学与工程(可...
- 37 篇 工商管理
30 篇 医学
25 篇 农学
18 篇 经济学
16 篇 法学
12 篇 教育学
11 篇 军事学
1 篇 文学
1 篇 艺术学

主题

35 篇 nuclear power pl...
24 篇 mathematical mod...
20 篇 computational mo...
20 篇 finite element m...
19 篇 optimization
17 篇 training
16 篇 analytical model...
16 篇 feature extracti...
15 篇 laboratories
14 篇 kinematics
13 篇 simulation
13 篇 algorithm design...
13 篇 data models
12 篇 monitoring
12 篇 predictive model...
12 篇 solid modeling
11 篇 reinforcement le...
11 篇 deep learning
11 篇 accuracy
10 篇 pressurized wate...

机构

410 篇 science and tech...
58 篇 tianjin key labo...
57 篇 science and tech...
57 篇 national demonst...
42 篇 national key lab...
39 篇 department of en...
36 篇 tsinghua nationa...
34 篇 school of softwa...
31 篇 nuclear power in...
30 篇 school of system...
30 篇 department of co...
27 篇 cas key laborato...
25 篇 nuclear power in...
22 篇 national key lab...
20 篇 school of comput...
17 篇 national demonst...
17 篇 institute of lau...
17 篇 key laboratory f...
15 篇 hefei national l...
15 篇 university of ch...

作者

39 篇 deng jian
23 篇 wang kan
23 篇 li qing
22 篇 yu hongxing
17 篇 qiu suizheng
16 篇 tian wenxi
15 篇 peng xingjie
14 篇 zhenyu zhang
14 篇 dan feng
14 篇 su guanghui
13 篇 chen zhi
13 篇 chen ping
13 篇 zhang xu
13 篇 yao dong
13 篇 li bin
12 篇 huang shanfang
12 篇 gong helin
12 篇 li jishun
12 篇 chai xiaoming
11 篇 yu yingrui

语言

1,849 篇 英文
70 篇 中文
52 篇 其他
3 篇 法文
1 篇 德文

检索条件"机构=National Key Laboratory of Science and Technology on Reactor System Design Technology"

共 1979 条记录，以下是191-200 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：

Investigation of turbulent Prt model and the segmentation rule for LBE turbulent heat transfer 18

Investigation of turbulent Prt model and the segmentation ru...

引用

18th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2019

作者： Li, Peiying Yu, Hongxing Li, Meifu Du, Sijia Yan, Mingyu Liu, Yu Zhu, Dahuan Zhang, He Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

With liquid metal like lead-bismuth alloy(LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt=A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental r esults, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism. © 2019 American Nuclear Society. All rights reserved.

关键词： Liquid metals

来源：评论

学校读者我要写书评

暂无评论

Testing Verification of Pressurizer Control in Nuclear Power Plant Based on Comparative Simulation 5th

Testing Verification of Pressurizer Control in Nuclear Power...

引用

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

作者： Zhang, Xu Deng, Zhi-Guang Peng, Hao Chen, Qi Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811634550

The pressurizer is an important equipment to maintain the stability of primary circuit in nuclear power plant. Pressurizer control system needs to be verified by effective testing. The most popular testing verification method on digital control system (DCS) is the discrete script tests. However it cannot make effective tests on the continuous variable process system. In this paper, a completed testing structure aiming on DCS is designed. The structure includes four levels: discrete script test, simulation test, stimulation test and online comparison test, which are in the progressive relationship, forming the comprehensive design and test system for the DCS. The system is applied to the verification of pressurizer in nuclear power plant. The HPR 1000 MW PWR model and pressurizer control system are used to make the simulation and verification tests. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

来源：评论

学校读者我要写书评

暂无评论

Influence of Different Factors on Fault Current Under DC Side Fault of Photovoltaic Grid Connected system 5

Influence of Different Factors on Fault Current Under DC Sid...

引用

5th IEEE Conference on Energy Internet and Energy system Integration, EI2 2021

作者： Luo, Yongyi Qing, Xianguo Peng, Renyong Huangfu, Yuzhao Cui, Ao Wang, Zhaosu Zhang, Zhiqiang Peng, Ziheng Liao, Chengyu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781665434256

Modular multilevel converter (MMQ has many advantages, such as low loss, flexible control and so on, which has been widely used in photovoltaic grid connected system. However, the photovoltaic output has some instability, and the damping of the flexible DC system is also low, which puts forward a severe test for the stable operation of the power grid. Firstly, this paper describes the basic structure of the photovoltaic system connected by MMC. Secondly, the photovoltaic grid connected system model based on PSCAD illustrates the fault current characteristics under DC side fault, and simulates and analyzes the influence of different factors in the system on the fault current. © 2021 IEEE

关键词： Power converters

来源：评论

学校读者我要写书评

暂无评论

Effects of core peripheral components melting on corium pool configuration in lower plenum of ACP1000 27

Effects of core peripheral components melting on corium pool...

引用

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

作者： Chen, Liang Yu, Hongxing Liu, Lili Dong, Huaping Deng, Jian Zhang, Hang Hou, Liqiang Wang, Xiaoji Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

The MAAP5 code and CFD method are adopted to estimate the melting behavior of core peripheral components for an Advanced China PWR, called ACP1000 after a large break accident to determine the amount of stainless steel relocating into the corium pool. The results suggest that larger portion of peripheral components melting in CFD method because of considering the radiation of fuel rods on shroud and barrel before core melting and ignoring the cooling of steam on the components in core. Effects of different configurations of corium pool on heat flux imposed on the vessel wall are also investigated in this work. The heat flux in light metal layer for the corium pool that obtained by MAAP5 code could be higher than that of CFD case due to its thinner light metal layer. Copyright © 2018-2019 by JSME

关键词： Heat flux

来源：评论

学校读者我要写书评

暂无评论

Simulation of esprit of hualong secondary passive residual heat removal system by using relap5 16

Simulation of esprit of hualong secondary passive residual h...

引用

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

作者： Feng, Li Science and Technology on Reactor System Design Technology Laboratory ChangShun Avenue 328 ShuangLiuCounty ChengDu China

ISBN: (纸本)9781510811843

A 3rd generation reactor called HuaLong is developed by China national Nuclear Corporation (CNNC). S econdary passive system (PRS) is designed to remove the residual heat during 72 hours after the reactor trips. To verify the ability of PRS system, the test facility ESPRIT is built in Chengdu which is 1:1 in heigh t and 1/62.5 in power. Tests are consisted by three Steady-states and one transient. One model of RELAP5 mod3.2 is developed and used to simulate the steady and transient tests. The results of RJELAP5 show con sistent with facility results, but the early phenomenon during transients is different between RELAP5 and tests as RELAP5 underestimates the direct-contact condensation power.

关键词： Condensation

来源：评论

学校读者我要写书评

暂无评论

Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast reactor 23rd

Atomic Scale Simulation on Liquid Metal Embrittlement Induce...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Bida, Zhu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811988981

Lead-Cooled fast reactor, which uses liquid lead or lead-bismuth alloy as coolants, is one of the six main reactors of the fourth generation reactor system. Most recently, the Lead-Cooled fast reactor received a lot of attention due to its safety, economy, and sustainability advantages. However, the liquid metal embrittlement (LME) could be a potential threat to the reliability of structure materials used in the Lead-Cooled fast reactor. Several experimental results show that the segregation of alloy element in grain boundary has a close association with liquid metal embrittlement. In this paper, by applying atomic simulation techniques, including molecular dynamics and Monte Carlo method, the segregation of lead element on poly-crystal ferritic and the resulting mechanical properties variations are simulated with the newest interatomic potential. The simulation results show that the lead element segregates on grain boundary. Moreover, such segregation of lead on grain boundary induced hardening and embrittlement of poly-crystal systems in the following uniaxial tensile testes. The atomic-scale mechanisms for such mechanical properties variation induced by lead element segregation are discussed carefully. In conclusion, these results verify that the segregation of lead element could be one of the critical mechanisms for LME, which well agree with previous experimental evidences. The present study may allow us to understand the mechanism of LME better. It can serve as a good reference for selecting more suitable materials for the Lead-Cooled fast reactor. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Grain boundaries

来源：评论

学校读者我要写书评

暂无评论

Effect of clamping failure on flow induced vibration and fretting wear of fuel rods

Effect of clamping failure on flow induced vibration and fre...

引用

ASME 2018 Pressure Vessels and Piping Conference, PVP 2018

作者： Huan-Huan, Qi Zhi-Peng, Feng Fu-Rui, Xiong Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791851654

Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, according to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. Next, by combining the correlation PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The clamping of fuel rod at various grids may fail due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the response amplitude of turbulent excitation at the bottom and top of the fuel rod is larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure over the rod wear depth depicts basically the same trend as on the maximum FIV amplitude. In addition to the FIV amplitude, wear depth is also related to rod natural frequency. By examining the multiplication of amplitude and frequency at the top and bottom grid, we found that the dimple failure has greater impact at the top grid. Copyright © 2018 ASME.

关键词： Natural frequencies

来源：评论

学校读者我要写书评

暂无评论

EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON PREDICTING TWO-PHASE FLOW key PARAMETERS OF RECTANGULAR CHANNEL 29

EXPLORATION AND APPLICATION OF ENSEMBLE LEARNING METHOD ON P...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Huang, Qingyu Yu, Yang Zeng, Hui Yang, Hui Zhang, Siyuan Yang, Kunlin Lin, Yuanfeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

Among the current mainstream nuclear reactor thermal-hydraulic calculation system analysis softwares (such as RELAP5, TRACE, etc.), different empirical or semi-empirical relationships are employed to calculate two-phase interfacial flow parameters. However, due to the lack of understanding of the physical laws of reactor fluids and the limited experimental data currently, the application scope and accuracy of these methods are relatively limited. As an important paradigm in machine learning methods, ensemble learning has received great attention in numerical prediction tasks in many fields due to its excellent robustness, nonlinear fitting ability and strong interpretability. In this study, three typical ensemble learning methods including random forest, gradient boosting regression tree, and extreme gradient boosting tree are used to build a two-phase interfacial parameter prediction model for rectangular channels. The investigated interfacial parameters include interfacial area concentration, void fraction, bubble velocity, bubble chord length, and bubble frequency. The result reveals that XGBoosting has the best performance in forecasting. In order to further improve the performance of the ensemble learning method and solve the problem of poor generalization ability of the homogeneous-ensemble model, the heterogeneous-ensemble strategy are adopted to combine different types of algorithms to provide a model with stronger scalability and generalization ability for two-phase flow interfacial parameter prediction. The results show that the usage of the heterogeneous-ensemble method can improve the prediction ability of the model to a certain extent, and further ensure the reliability of the data-driven model for the prediction of the two-phase interfacial flow key parameters in the rectangular channel. Copyright © 2022 by ASME.

关键词： Two phase flow

来源：评论

学校读者我要写书评

暂无评论

STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCLEAR reactor UNDER OCEAN CONDITION 29

STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCL...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Cheng, Kun Chu, Xiao Qiu, Zhifang Xian, Lin Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886380

Different from land-based reactors, floating reactors are affected by the ocean condition. Numerical simulation is conducted to study the effects of heaving and rolling motion on typical floating nuclear reactor (FNR) under rated power operation by ARSAC code with ocean condition calculation module. The calculation results have shown that heaving motion can cause the periodic fluctuations in the flow rate of both the primary and secondary loops in FNR. The amplitude of heaving motion is the main influence factor on the flow rate fluctuation of the primary loop while the effect of heaving motion period is limited. Rolling motion can cause the out-phase flow fluctuation in the two symmetric loops of the primary system. However, the rolling motion has limited impact on the core flow rate and cannot cause dramatic changes of system parameters at the core. Within the range of motion parameters used in this paper, heaving and rolling motion cannot have a substantial impact on the safe operation of FNR under rated power operation. Copyright © 2022 by ASME.

关键词： Oceanography

来源：评论

学校读者我要写书评

暂无评论

NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHARACTERISTICS IN LBE-COOLED HELICAL CRUCIFORM FUEL AND WIRE-WRAPPED FUEL 31

NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHA...

引用

2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Qi, Zhang Haoyu, Wang Cao, Junxian Yuanming, Li Chenxi, Li Zhenhai, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791888261

The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surrounds the cylindrical fuel is developed and promoted, which may be accept by the LBE-cooled fast reactor. The helical cruciform fuel of the Lightbridge corporation has similar geometrical features with the wire-wrapped fuel. The cross-section of helical cruciform fuel consists of the convex blades and the concave elbows, and the cross-section is rotated around the axis. The blade and elbow of helical cruciform fuel function as the wires while they are part of the nuclear fuel. In the present work, the comparison on the thermal-hydraulic and flow-induced vibration (FIV) characteristics of the helical cruciform fuel and wire-wrapped fuel is carried out by numerical simulation. The fluid-structure interaction is solved by the coupling between the fluid and solid domains, and the transient turbulence fluctuation is solved by large eddy simulation (LES). The results show that the convective heat transfer is enhanced by the spiral structures even though the hot spot phenomenon is caused simultaneously. The hot spot temperature of the helical cruciform fuel is larger than the wire-wrapped fuel duo to the non-uniform distributed heat flux. As the coolant in the flow channel is blocked by the blades and wires, the directional transverse flow is generated and thus the transverse mixing is enhanced. The one-way mixing with opposite direction occurs in the two sides of the single rod. There are similarities in the FIV characteristics for the spiral shaped fuels. The maximum vibration displacement and dominant vibration frequency of the helical cruciform fuel and wire-wrapped fuel are almost the same. © 2024 by ASME.

关键词： Fluid structure interaction

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198页 << < 16 17 18 19 20 21 22 23 24 25 > >>

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：