• KSTLE International Journal
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• v.10 no.1_2
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• pp.33-36
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• 2009

The nuclear steam generator is composed of a bundle of tubes. The length of these tubes is very long, but their diameter is small. Fluid exists inside of the steam generator and its flow causes vibration, therefore these tubes are supported by anti-vibration bars. The wear damage due to the vibration is known as fretting wear, which should be minimized to ensure the safety of the plants. Research needs to be done about decreasing the amount of fretting wear. Hard coatings have proven to be very effective in reducing the amount of wear. The commercial coatings of TiN and CrN have excellent wear resistance and are used to protect the Inconel tube from fretting wear. The tube-on-flat type tester was used for conducting the fretting wear tests. It was found that the wear amounts of the coated tubes decreased depending on the coating thickness. CrN was found to be very effective in reducing the wear, while the wear amounts were dependent on the coating thickness in the case of TiN and a thick coating of TiN was very effective on wear resistance.

• Proceedings of the IEEK Conference
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• 1998.06a
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• pp.223-226
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• 1998

The control objective of steam generator water level in the secondary circuit of a nuclear power plant is to regulate the water level at the desired set point. The dynamics of steam generators is non-linear in nature. The task of modelling such plant is very difficult and especially so when plant operating conditions change frequently. In these reasons, conventional PI gains over all pover range will not work efficiently and a manual control is generally used in low power operation. Therefore the robust H$\infty$ controller design method should be required. In this paper, we design the robust H$\infty$ controller for water level control of steam generator using a mixed H$\infty$ optimization with model-matching method. Firstly we choose the desired model that has good disturbance rejection performance. Secondly we design a stabilizing controller to keep the model-matching error small and also provide sufficiently large stability margin against additive perturbations of the nominal plant.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.29-33
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• 2008

As the measures of nuclear power plant utilities and manufacturers to reduce the defects of tube expansion region during manufacturing steam generators, many types of NDEs(Non-Destructive Examinations) are conducted to inspect the expansion region. The expansion region of tube is subject to degrade because of stress concentration induced by tube expansion, sludge pile and high temperature. So the inspections for tube expansion region have been reinforced. Liquid penetrant test, helium leak test, Bobbin profile test and hydraulic test are performed to confirm the integrity of tube expanded by hydraulic expansion method. Liquid penetrant test and helium leak test are used to inspect seal weld region on tubesheet end part. Bobbin Profile test is used to inspect fully the expanded region of steam generator tube. Hydraulic test finally verifies the integrity of seal weld region on tubesheet end part.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.1-8
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• 2010

Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.869-870
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.943-944
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• 2005

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1677-1688
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• 2020

In the sodium-cooled fast reactor, the steam generator is a heat exchange device between sodium and water, which may cause leakage, resulting in a sodium-water reaction accident, which in turn affects the safe operation of the entire nuclear reactor. To this end, the electromagnetic vortex flowmeter is used to detect leakage of the steam generator and its signal processing method is studied in this paper. The hydraulic experiment was carried out by using water instead of liquid sodium, and the sensor output signal of the electromagnetic vortex flowmeter under different gas injection volumes was collected. The bubble noise signal is reflected by the base line of the sensor output signal. According to the relationship between the proportion of the bubble noise signal in the sensor output signal and the gas injection volume, a signal processing method based on the energy ratio calculation is proposed to detect whether the water contains bubbles. The gas injection experiment of liquid sodium was conducted to verify the effectiveness of the signal processing method in the detection of bubbles in sodium, and the minimum detectable leak rate of water in the steam generator was detected to be 0.2 g/s.

• Proceedings of the KIEE Conference
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• 1995.11a
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• pp.172-174
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• 1995

About a half of Electric power is generated by nuclear power plants in korea. So, the stable operation of nuclear power plant is very important for suppling the essential national electric power. A S/G(Steam Generator) level control is the most difficult system in PWR(Pressurized Water Reactor) nuclear power plant. Because of the non-linear and the non-nominal response of S/G level control, it Is very difficult to control the level by automatic mode or manual mode. The goal of this study is to establish and verify a advanced control algorithm by analyzing, modelling, stability calculation, controller parameter calculation, simulation for S/G level control system.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.155-164
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• 1995

The steam generator water level control system is studied as a class of randomly sampled nonlinear control systems. The sampling interval and the loop amplification factor are considered as random variables in order to take the operator behavior in account. Stochastic stability using Lyapunov method is used without determining such Lyapunov function. The derived stability criterion is verified with time-domain simulation using the data of CANDU type nuclear power plant, Wolsung 1.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.459-468
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• 2013

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.