scholarly journals Steam Generator Strength Computation for Nuclear Electric Power Plants of a Low Power

2021 ◽  
pp. 24-35
Author(s):  
Volodymyr Kravchenko ◽  
Oleksandr Lymarenko ◽  
Zhou Xiaolong ◽  
Kiril Khromiy ◽  
Yehor Buchka
Keyword(s):  
Heat Exchange ◽  
Low Power ◽  
Steam Generator ◽  
Heat Carrier ◽  
Nuclear Power ◽  
Power Plants ◽  
Branch Pipe ◽  
Cylindrical Part ◽  
The Body ◽  
External Branch

Today, the world oversees an explosive development of the nuclear power stations (NPS) of a low power. Most projects deal with pressurized water reactors and as a matter of fact with steam generators (SG). Ukraine has a well-developed engineering industry backbone that can be used for the production of the equipment required for the nuclear power plants of a low power. This scientific paper delves into the computations of the strength of elements used for the monotube steam generator with cylindrical coils that is the most presentable of all the projects in question in IAEA materials. Appropriate methods were developed to perform structural computations and steam generator strength computations. The mathematical model was developed that allows us to perform strength computations of the SG elements making use of the analytical method with reference to the Regulations and do simulations using the ANSYS software code. The specified elements include the body elements, in particular the cylindrical part, the flange, the bottom and the cover, including the heat carrier branch pipe and heat exchange tubes. The comparison of the data obtained by both methods showed their similarity and accordingly, the accuracy of the data that are indicative of the need for an increase in the wall thickness of the cylindrical part of the external branch pipe intended for the heat carrier. The body bottom strain for calculated dimensions exceeds the permissible value by 1.56 %. Since this value is 5 % lower than permissible values it is deemed that the strength condition is passed through. The simulation proved that the strength conditions are met for heat exchange tubes, for the body, the body cover, the body flange, the conical part of the external branch pipe intended for the heat carrier. Based on the analysis done, we would like to recommend performing strength computations using the normative method with the subsequent check out by the simulation using the computer code.

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.

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

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