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

共 626 条记录，以下是291-300 订阅

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High-accuracy time synchronization method in highly reliable fiber-channel communication system 27

High-accuracy time synchronization method in highly reliable...

引用

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

作者： Chen, Qi Jiang, Wei Tu, Xiaodong Yu, Bo Liu, Mingxing Zhang, Xu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China School of Information and Communication Engineering University of Electronic Science and Technology of China Chengdu Sichuan China

ISBN: (纸本)9784888983051

Accompanied with strict requirements on accuracy and determinacy in nuclear power industrial safety Digital Control system (DCS), time-triggered technology has been used in Ethernet widely, while another network protocol, Fiber-Channel (FC) protocol which is popular in avionic electronic systems with hyper reliability in port to port communication stands still, even though the optical fiber cable is quite adopted by other protocols like Ethernet. One important reason is the lack of the capability to handling time sensitive tasks in FC protocol itself. In order to bring Fiber-Channel protocol into real-time applications, like safety DCS for nuclear industry, with the situation that there are pretty few researches on the time synchronization in FC protocol, this paper makes fundamental research on the solution to achieve high-accuracy time synchronization in FC protocol networks, which has been proved a vital important step to realize real time communication according to Time-Triggered Ethernet (TTE)(H. Kopetz al, 2005). By analyzing current Fiber-Channel network time communication strategies and demo-boards testing, this paper designs a method to achieve the same level or better synchronization accuracy in FC network with that in TTE, which figures out a new direction in this blank field of time synchronization in FC protocol. Furthermore, this method only asks for few resources and slight modifications on current communications for future development and improvement. Copyright © 2018-2019 by JSME

关键词： Accident prevention

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A Transformer-based Time Series Prediction Model with Contrastive Learning for Nuclear Power system State Monitoring 15

A Transformer-based Time Series Prediction Model with Contra...

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15th IEEE Global Reliability and Prognostics and Health Management Conference, PHM-Beijing 2024

作者： Gou, Pengfei Zheng, Daiwei Tang, Lei Liu, Mengjuan School of Information and Software Engineering University of Electronic Science and Technology of China UESTC Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China NPIC Chengdu China

ISBN: (纸本)9798350354010

Prognostics and Health Management (PHM) is currently the mainstream solution for the condition monitoring and maintenance of industrial systems. In PHM, the system condition is monitored by monitoring the operational status parameters of the system to diagnose equipment faults, predict the remaining useful life, etc. Utilizing sequential operational status parameters to complete prediction and classification tasks is usually modeled as a time series processing task. Therefore, a high-performance time series prediction model is crucial for PHM. This paper focuses on the use of time series prediction models to achieve the prediction of key state parameters of the nuclear power system. However, there are still some challenges when directly applying existing time series prediction models to industrial systems. First, obtaining operational data on the nuclear power system is relatively difficult, resulting in a small amount of training data, which causes many models to face the risk of insufficient training or overfitting. Second, most of the existing time series prediction models are composed of high- and low-frequency decoupling modules, while the operational data of nuclear power systems often have fewer high-frequency components, making it difficult for many existing SOTA models to achieve good results. In order to address these challenges, we have developed a Transformer-based time series prediction model with contrastive learning for monitoring the status of nuclear power systems. We evaluated the new model on three nuclear power simulation datasets and four power transformer datasets. The experimental results show that the new model's performance is superior to baseline models. We also proved the effectiveness of our improvement through ablation experiments. © 2024 IEEE.

关键词： Prediction models

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Research about the influence of passive residual heat removal system on SGTR accident 27

Research about the influence of passive residual heat remova...

引用

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

作者： Wu, Yirui Shan, Jianqiang Xu, Tao Ma, Haifu Yang, Junyun Yu, Hong Wang, Junlong School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an Shaanxi Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan Province China

ISBN: (纸本)9784888983051

Since the passive safety system turns out to be a desirable method to achieve simplification and increase the reliability of the performance of essential safety functions, passive technology with inherent safety features has been generally applied to the design of advanced nuclear power plant. The PWR (pressurized water reactor) NPP (nuclear power plant) in this paper is a kind of mature pressurized water plants in China, and is proposed to be improved with some passive systems to enhance safety performance. The proposed and adapted passive safety systems for this PWR power plant in this paper contain: (1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);(5) the PRHRS (passive residual heat removal system), which is installed on the reactor coolant system (RCS) loop. In this research, the behaviors of passive safety systems are studied and transient characteristics of the primary and secondary side of the reactor during the Steam Generator Tube Rupture (SGTR) accident are investigated using the best-estimate transient simulation code Relap5/MOD3.3. Results reported in this paper indicate that making use of passive safety systems especially the PRHRS could ensure reactor safety during the SGTR accident. Copyright © 2018-2019 by JSME

关键词： Pressurized water reactors

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Interfacial area transport model for bubbly flow system in vertical rod bundle with spacer grids

Interfacial area transport model for bubbly flow system in v...

引用

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

作者： Liu, Hang Lai, Jianyong Li, Yi Zhang, Yulong Liu, Minghao Pan, Liangming Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Nuclear Engineering Chongqing University Chongqing400044 China

ISBN: (纸本)9780791883785

In relation to the modeling of the one-dimensional interfacial area transport equation for an adiabatic bubbly flow in a vertical rod bundle, some existing models of sink and source terms were reviewed and evaluated. Based on the reviewed interaction mechanisms of bubbles and turbulent eddies, a new interfacial area transport model has been proposed. Two important impacts on bubble interaction have been taken into account in the new model: The effects of spacer grids with mixing vanes and the impacts of geometry structure. The spacer grids breakup large bubbles into small bubbles resulting in enhanced bubble random collision at the downstream of the spacer grids. Void transport is the main contribution between spacer grids. The new interfacial area transport model has been evaluated against the obtained experimental data in 17 bubbly flow conditions. The results indicate that the new model can predict the interfacial area concentration with the relative error of 19.9%. It is recognized that the proposed model is promising for predicting the interfacial area concentration for a bubbly flow in a rod bundle. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Nuclear engineering

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Comparison of SGTR accident with two different passive residual heat removal systems 27

Comparison of SGTR accident with two different passive resid...

引用

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

作者： Wu, Yirui Shan, Jianqiang Xu, Tao Ma, Haifu Yang, Junyun Yu, Hong Wang, Junlong Deng, Yaxin School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an Shaanxi Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan Province China

ISBN: (纸本)9784888983051

Since the passive safety system turns out to be a desirable method to achieve simplification and increase the reliability of the performance of essential safety functions, passive technology has been generally applied to the design of advanced nuclear power plant. The traditional generation II+ nuclearpower plant is proposed to be improved with some passive systems to enhance safety performance. There are two series of passive safety systems for this kind of nuclear power plant in this paper, one of them contains: (1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);and (5) the PRHRS (passive residual heat removal system), which is installed on the reactor coolant system (RCS) loop;the other contains:(1) the RMT (reactor makeup tank);(2) the A-ACC (advanced accumulators);(3) the IRWST (in-containment refueling water storage tank);(4) the PDS (passive depressurization system);(5) the PEFS(passive emergency feed water system), which is installed on the secondary side of SGs. In this research, the transient characteristics of SGTR accident with the two different passive residual heat removal systems which are PRHRS and PEFS is investigated using the best-estimate transient simulation code Relap5/MOD3.3. The calculation results show that both of PRHRS and PEFS could remove the decay heat from the primary loop efficiently to ensure the reactor safety and reduce the radioactive leakage to mitigate the accident impact during the SGTR accident. Due to the significant cooling effect of PRHRS and PEFS, the water level of broken SG would not be overfilled in both conditions, which is different with the SGTR accident of traditionalgeneration II+nuclear power plant. Compared with the PEFS, the PRHRS performed better in cooling efficiency, long-term cooling performance, reducing radioactive leakage capacity and mitigating accident effect. Copyright © 2018-2

关键词： Nuclear power plants

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The inexact gauss-newton method for core power distribution monitoring in pressurized water reactors

The inexact gauss-newton method for core power distribution ...

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2013 Winter Meeting on Transactions and Embedded Topical Meetings: Risk Management for Complex Socio-Technical systems, 2nd ANS SMR 2013 Conference, Nuclear Nonproliferation - 1st Fission to the Future

作者： Peng, Xingjie Wang, Kan Li, Qing Department of Engineering Physics Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610041 China

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Study on optimization design for CSR1000 core

Study on optimization design for CSR1000 core

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作者： Wang, Lianjie Yang, Ping Lu, Di Zhao, Wenbo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No. 328 Section 1 Changshun Avenue Shuangliu County Chengdu610213 China

An optimization design of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWe (CSR1000) conceptual core is proposed. Steady-state performance of the proposed core is then studied with the SCWR core steady-state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature, and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path, which is suited with assembly power, is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design. Copyright © 2018 by ASME.

关键词： Water cooled reactors

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Radial burnup profile fitting of a high burnup pellet

Radial burnup profile fitting of a high burnup pellet

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2013 ANS Annual Meeting - Next Generation Nuclear Energy: Prospects and Challenges

作者： Gao, Lijun Jiang, Shengyao Chen, Bingde Department of Engineering Physics Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610041 China

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Partheno-genetic algorithm approach for optimal in-core neutron detector location

Partheno-genetic algorithm approach for optimal in-core neut...

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2013 ANS Annual Meeting - Next Generation Nuclear Energy: Prospects and Challenges

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Validation of application of dislocation punching condition to estimating the HBS pore pressure

Validation of application of dislocation punching condition ...

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2014 Annual Meeting on Transactions of the American Nuclear Society and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear reactors, NSFM 2014

作者： Gao, Lijun Chen, Bingde Jiang, Shengyao Department of Engineering Physics Tsinghua University Beijing 100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610041 China

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