• Proceedings of the KIEE Conference
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• 2009.07a
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• pp.1678_1679
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• 2009

원전에 설치되어 있는 노외 핵계측 검출기는 설계수명과 품질보증 수명의 차이가 크기 때문에 적절한 검출기 교체 프로그램의 수립이 어렵다. 따라서 본 논문에서는 노외 핵계측검출기에 대한 노화 진행정도 판단 및 최적 교체주기 수립을 위한 원전 노외 핵계측 검출기 성능진단 시스템 및 분석기술에 대해 고찰하였다.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.2036-2037
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• 2008

An Emergency Diesel Generator (EDG) manufactured by a French company Wartsila SACM, is a tandem type engine, consisted of two 10 cylindered diesel engines on each side. Manual provided by the manufacturer states that engine bearing requires inspection every 15 years. However, it is difficult for an inspector to access through a manhole located in the lower compartment of engine. Furthermore, during a routine or scheduled maintenance, it is not possible to disassemble main engine bearing and crank shaft, and perform inspection. Two methodologies are suggested here to resolve the problem. One method is to lift the engine and partially perform the maintenance service, and the other method is to disassemble the engine completely and to perform maintenance service by the manufacturer. Pros and cons of two methodologies were thoroughly compared.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.224-233
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• 2017

In Korea, attempts have been made to develop high strength concrete for the safety and design life improvement of nuclear power plants. In this study, the cracking potentials of nuclear power plant-high strength concretes (NPP-HSCs) containing industrial by-products with W/B 0.34 and W/B 0.28, which are being reviewed for their application in the construction of containment structures, were evaluated through autogenous shrinkage, unrestrained drying shrinkage, and restrained drying shrinkage experiments. The cracking potentials of the NPP-HSCs with W/B 0.34 and W/B 0.28 were in the order of 0.34FA25 > 0.34FA25BFS25 > 0.34BFS50 > 0.34BFS65SF5 and 0.28FA25SF5 >> 0.28BFS65SF5 > 0.28BFS45SF5 > 0.28 FA20BFS25SF5, respectively. The cracking potentials of the seven mix proportions excluding 0.28FA25SF5 were lower than that of the existing nuclear power plant concrete; thus, the durability of a nuclear power plant against shrinkage cracking could be improved by applying the seven mix proportions with low cracking potentials.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2576-2583
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• 2024

In this paper, the similarity criterion and dimensionless conversion method combined with the elasticity condition and Hooke's law are used to derive the functional relationship of the maximum effective value of the vibration velocity between the prototype pump and the model pump. The seawater circulation pump of a nuclear power plant is used as the prototype pump, and the model pump is obtained by performance conversion and choosing the appropriate scale, and the vibration state of the model pump under different flow rates is measured and analyzed. The vibration data of the model pump through the function relationship to find out the vibration parameters of the prototype model pump, and compare with the vibration data of the seawater circulation pump in reality. It can be seen that with the increase of flow rate, the maximum effective value of the vibration velocity of both model and prototype decreases and then increases, and the relative error is small, the maximum value is 7.7757%. Therefore, it can be considered that the functional relationship of model pump converted to prototype pump derived in this paper can be used to analyze the vibration of the actual seawater circulation pump of coastal nuclear power plant.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.897-907
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• 2020

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.443-444
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.425-426
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.207-208
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.453-454
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• 2004