• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.1
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• pp.12-17
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• 2006

A steam generator of nuclear power plant has thousands of thin tubes. These tubes play an important role in maintaining the pressure boundary between the primary and secondary side of nuclear power plant. The steam generator tube is easy to be damaged because of the severe operating conditions such as the high temperature and pressure. Therefore, tremendous efforts are made to assess the structural integrity of the steam generator tubes. The eddy current test is the most popular non-destructive test to assess the integrity of the tubes. However, the eddy current test has the limitation to size the flaw accurately because the eddy current signal behavior depends on the total volume of flaw. This paper shows the possibility that the ultrasonic test method can be applied to detect the flaws in the steam generator tubes and to measure them quantitatively. From the test results, it is expected that if the ultrasonic test is put to practical use in the steam generator tube inspection, the inspection results will be improved.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.134.3-134
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• 2001

The dynamic behavior of heat recovery steam generators for combined cycle power plant is simulated in cases of startup and shutdown conditions. To ensure performance and design data, dynamic model of the HRSG was developed and dynamic simulation was performed. The dynamic analysis will undoubtedly reduce costs which is associated with plant startup and contribute to a smooth commercial plant operation.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.179-190
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• 2003

The multiple steam generator tube rupture (SGTR) event scenario with available safety systems was experimentally and analytically evaluated. The experiment was conducted on the large scaled test facility to simulate the multiple SGTR event and investigate the effectiveness of operator actions. As a result, it indicated that the opening of pressurizer power operated relief valve was significantly effective in quickly terminating the primary-to-secondary break flow even for the 6.5 tubes rupture. In the analysis, the recent version of RELAP5 code was assessed with the test data. It indicated that the calculations agreed well with the measured data and that the plant responses such as the water level and relief valve cycling in the damaged steam generator were reasonably predicted. Finally, sensitivity study on the number of ruptured tubes up to 10 tubes was performed to investigate the coolant release into atmosphere. It indicated that the integrated steam mass released was not significantly varied with the number of ruptured tubes although the damaged steam generator was overfilled for more than 3 tubes rupture. These findings are expected to provide useful information in understanding and evaluating the plant ability to mitigate the consequence of multiple SGTR event.

• Journal of Institute of Control, Robotics and Systems
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• v.4 no.3
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• pp.321-328
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• 1998

When we control the level of the steam generator in the nuclear power plants, a swell and shrink arises from many disturbances such as feed water rate, feed water temperature, main steam flow rate, and coolant temperature. If we use the conventional type of PI controller in this system, we will not have stability during controlling at lower power, the removal function of disturbances, and a load follow-up control effectively. In this paper, we study the application of a 2-Degree of Freedom(2-DOF) PID controller to the level control of the steam. generator of nuclear power plants through the simulation and the experimental steam generator. We use the parameters $\alpha$, $\beta$, $\gamma$ of the 2-DOF PID controller for the removal of disturbances and the parameters Kp,Ti,Td of the conventional type of PID controller for controlling setpoint. The back-propagation learning algorithm of neural network is used for tuning the 2-DOF PID controller. We can find satisfactory results of the removal of the disturbances and the tracking function in the change of setpoint through the simulation and experimental steam generator.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.20 no.6
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• pp.416-426
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• 2008

The boiler tube failure often experienced in the superheater of a utility boiler can seriously affect the economic and safe operation of the power plant. It has been known that this failure is mainly caused by the thermal load deviation in the superheater tube system, and deeply intensified by the non-uniform distribution of steam flow rates. The nonuniform steam flow is distinctively prominent at low power load rather than at full power load. In this paper, we analyze the steam flow distribution in the superheater tube system by using one dimensional flow network model. At 30% power load, the deviation of steam flow rate is predicted to be within 0.8% of the averaged flow rate. This deviation can be reduced to 0.1% and 0.07% by assuming two cases, that is, the removal of 13th tube at each tube rows and the installation of intermediate header, respectively. The assumed two cases would be effective for the uniform steam flow distribution across 85 superheater tube rows.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2003.10a
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• pp.39-42
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• 2003

Steam Generator has been manufactured by welding process after partial manufacturing of various steel forgings such as shell, head and tube sheet. Usually, these steel forgings are made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced open die forging development status of the large steel forgings which is used for the steam generator of 1,400MW next generation nuclear power plant.

• Proceedings of the Korean Institute of Intelligent Systems Conference
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• 2002.05a
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• pp.203-206
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• 2002

This paper describes fuzzy control methodologies of the steam generator which have nonlinear characteristics in the nuclear power plant. Actually, the steam generator part of the power generator has a problem to control water level because it has complex components and nonlinear characteristics. In order to control nonlinear terms of the model, Takagj-Sugeno (75) fuzzy system is used to design a controller. In designing procedure, intelligent digital redesign method is used to control the nonlinear system. This digital controller keeps the performance of the analog controller. Simulation examples are included for ensuring the proposed control method.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.24 no.4
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• pp.116-122
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• 2010

This paper describes the up-grade of the turbine governor for steam turbine due to its poor operation from long time use. The analog type governor of the unit 1 in Kori nuclear power plant in Korea was removed and the new digital type turbine governor was developed and installed. The procedure for the actual application, site adaptability test using dynamic simulator and the result of actual operation are described here. And the program for nuclear steam turbine is suggested here.

• Proceedings of the KIEE Conference
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• 2006.10c
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• pp.591-593
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• 2006

The differential pressure method has been used in the level measurement of steam generator in nuclear power plant. Two sensing lines from a steam generator to a pressure transmitter are needed to measure the high pressure and low pressure. The fluid conditions in the sensing line require the uniform phase with no bubbles and the slope of sensing line should be installed with forward slope. The expansion of the bubble volume according to the upper pressure and the reverse slope of sensing lines explain how the level errors took place.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10b
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• pp.601-604
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• 1996

It is well known that steam generator water level control at low power operation has many difficulties in a PWR (pressurized water reactor) nuclear power plant. The reverse process responses known as shrink and swell effects make it difficult to control the steam generator water level at low power. A new automatic control logic to remove the reverse process responses is proposed in this paper. It is implemented in PLC (programmable logic controller) and evaluated by using test equipment in Korea Atomic Energy Research Institute. The simulation test shows that the performance requirements is met at low power (below 15%). The water level control by new control logic is stabilized within 1% fluctuation from setpoint, while the water level by YGN 3 and 4 control logic is unstable with the periodic fluctuation of 25% magnitude at 5% power.