scholarly journals Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

2017 ◽  
Vol 2017 ◽  
pp. 1-8
Author(s):  
Jianghai Li ◽  
Jia Meng ◽  
Xiaojing Kang ◽  
Zhenhai Long ◽  
Xiaojin Huang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Heterogeneous Data ◽  
Wireless Sensor ◽  
Support Vector ◽  
Intelligent Monitoring ◽  
Novel Algorithms ◽  
Non Gaussian ◽  
High Temperature Gas

High-temperature gas-cooled reactors (HTGR) can incorporate wireless sensor network (WSN) technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs). This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA) and support vector machine (SVM) are developed.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

The Application of System CFD to the Design and Optimization of High-Temperature Gas-Cooled Nuclear Power Plants

2008 ◽  
Vol 130 (3) ◽  
Author(s):  
Gideon P. Greyvenstein
Keyword(s):  
Steady State ◽  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Pressure ◽  
Critical Parameters ◽  
Wide Range ◽  
Component Models ◽  
High Temperature Gas

The objective of this paper is to model the steady-state and dynamic operation of a pebble-bed-type high temperature gas-cooled reactor power plant using a system computational fluid dynamics (CFD) approach. System CFD codes are 1D network codes with embedded 2D or even 3D discretized component models that provide a good balance between accuracy and speed. In the method presented in this paper, valves, orifices, compressors, and turbines are modeled as lumped or 0D components, whereas pipes and heat exchangers are modeled as 1D discretized components. The reactor is modeled as 2D discretized system. A point kinetics neutronic model will predict the heat release in the reactor. Firstly, the layout of the power conversion system is discussed together with the major plant parameters. This is followed by a high level description of the system CFD approach together with a description of the various component models. The code is used to model the steady-state operation of the system. The results are verified by comparing them with detailed cycle analysis calculations performed with another code. The model is then used to predict the net power delivered to the shaft over a wide range of speeds from zero to full speed. This information is used to specify parameters for a proportional-integral-derivative controller that senses the speed of the power turbine and adjusts the generator power during the startup of the plant. The generator initially acts as a motor that drives the shaft and then changes over to a generator load that approaches the design point value as the speed of the shaft approaches the design speed. A full startup simulation is done to demonstrate the behavior of the plant during startup. This example demonstrates the application of a system CFD code to test control strategies. A load rejection example is considered where the generator load is suddenly dropped to zero from a full load condition. A controller senses the speed of the low pressure compressor/low pressure turbine shaft and then adjusts the opening of a bypass valve to keep the speed of the shaft constant at 60rps. The example demonstrates how detailed information on critical parameters such as turbine and reactor inlet temperatures, maximum fuel temperature, and compressor surge margin can be obtained during operating transients. System CFD codes is a powerful design tool that is indispensable in the design of complex power systems such as gas-cooled nuclear power plants.

scholarly journals Comparison Analysis of Selected Nuclear Power Plants Supplied With Helium from High-Temperature Gas-Cooled Reactor

2018 ◽  
Vol 25 (s1) ◽  
pp. 204-210 ◽  
Author(s):  
Natalia Szewczuk-Krypa ◽  
Marta Drosińska-Komor ◽  
Jerzy Głuch ◽  
Łukasz Breńkacz
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Social Aspect ◽  
Passive Safety Systems ◽  
System A ◽  
Safety Systems ◽  
High Temperature Gas

Abstract The article presents results of efficiency calculations for two 560 MW nuclear cycles with high-temperature gas-cooled reactor (HTGR). An assumption was made that systems of this type can be used in so-called marine nuclear power plants. The first analysed system is the nuclear steam power plant. For the steam cycle, the efficiency calculations were performed with the code DIAGAR, which is dedicated for analysing this type of systems. The other system is the power plant with gas turbine, in which the combustion chamber has been replaced with the HTGR. For this system, a number of calculations were also performed to assess its efficiency. Moreover, the article names factors in favour of floating nuclear power plants with HTGRs, which, due to passive safety systems, are exposed to much smaller risk of breakdown than other types of reactors which were in common use in the past. Along with safety aspects, it is also economic and social aspect which make the use of this type of systems advisable.

Real-Time Simulation Platform for the Design and Verification of the Operation Strategy of the HTR-PM

2013 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang
Keyword(s):  
High Temperature ◽  
Real Time ◽  
Nuclear Power ◽  
Power Plants ◽  
Simulation Platform ◽  
Operation Strategy ◽  
Real Time Simulation ◽  
Time Simulation ◽  
Successful Design ◽  
High Temperature Gas

Because of the inherent safety feature, the modular high temperature gas-cooled reactor (MHTGR) has been seen as one of the best candidates for the next generation of nuclear power plants. Power-regulation is a key technique to provide safe, stable and efficient plant operation for the multi-reactor-module power plants. Based on the successful design and operation of Chinese test high temperature gas-cooled reactor HTR-10, the high temperature gas cooled reactor pebble-bed module (HTR-PM) project is proposed. Since the HTR-PM is composed of two interconnected nuclear steam supply system (NSSS), the operation strategy of the multi-reactor-modular HTR-PM is different from that of those one-reactor-modular high temperature reactors such as HTR-10. In order to design and verify the operation strategy for HTR-PM, it is necessary to establish a real-time simulation platform. The design and development of this platform is presented in this paper.

scholarly journals Patch-Type Vibration Visualization (PVV) Sensor System Based on Triboelectric Effect

Sensors ◽  
2021 ◽  
Vol 21 (12) ◽  
pp. 3976
Author(s):  
Sun Jin Kim ◽  
Myeong-Lok Seol ◽  
Byun-Young Chung ◽  
Dae-Sic Jang ◽  
Jonghwan Kim ◽  
...  
Keyword(s):  
Signal Processing ◽  
Low Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Computing System ◽  
Sensor System ◽  
Wireless Sensor ◽  
Sensor Systems ◽  
Patch Type ◽  
Triboelectric Effect

Self-powered wireless sensor systems have emerged as an important topic for condition monitoring in nuclear power plants. However, commercial wireless sensor systems still cannot be fully self-sustainable due to the high power consumption caused by excessive signal processing in a mini-electronic computing system. In this sense, it is essential not only to integrate the sensor system with energy-harvesting devices but also to develop simple data processing methods for low power schemes. In this paper, we report a patch-type vibration visualization (PVV) sensor system based on the triboelectric effect and a visualization technique for self-sustainable operation. The PVV sensor system composed of a polyethylene terephthalate (PET)/Al/LCD screen directly converts the triboelectric signal into an informative black pattern on the LCD screen without excessive signal processing, enabling extremely low power operation. In addition, a proposed image processing method reconverts the black patterns to frequency and acceleration values through a remote-control camera. With these simple signal-to-pattern conversion and pattern-to-data reconversion techniques, a vibration visualization sensor network has successfully been demonstrated.

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