• Nuclear Engineering and Technology
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• 제41권3호
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• pp.287-294
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• 2009

Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.

• Elastomers and Composites
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• 제55권3호
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• pp.222-228
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• 2020

Rubber materials, used in nuclear power plants, need high heat-oxidation resistance to curing or cracking under a heat aging environment. This is because they are applied to environments with high temperature, high humidity, and radiation exposure. Nuclear radiation causes additional hardening or degradation, therefore, rubber materials need radiation resistance that satisfies the general and any accidental conditions produced in the power plant. Therefore, in this study, we developed a rubber material with excellent heat and radiation resistance for the diaphragm seal of a nuclear steam generator nozzle dam. The rubber material greatly improved the reliability of the steam generator nozzle dam. In addition, 30 inch and 42 inch diaphragm seals were manufactured using the developed rubber material. A nozzle dam was installed in a nuclear power plant and tested under the same conditions as a steam generator to evaluate safety and reliability. In the future, the performance and safety of diaphragm seals developed through field tests of nuclear power plants will be evaluated and applied to currently operating and new nuclear power plants.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2000년도 제15차 학술회의논문집
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• pp.359-359
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• 2000

Generally, level control in the steam generator of a nuclear power plant is difficulty process control, because the low power operating can lead nonminimum phase characteristics(swell and shrink phenomenon) and flow measurement are unreliable and nonlinear characteristics. This paper presents a framework for solving this problem based on the constrained linear model predictive control and introduces the design of method for the level of the controller in the entire operating power of the steam generator, and compares with conventional PI controller.

• Journal of the Korean Society of Safety
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• 제19권2호
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• pp.11-15
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• 2004

The fretting-fatigue by leaking is one of the significant degradation in steam generator tubes. In this study, the burst pressure of inclined damaged steam generator tubes were obtained from three criterions by using the finite element method. The analysis results were also compared with the experiment data from published references and they showed a good agreement with the experiment data.

• Nuclear Engineering and Technology
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• 제35권3호
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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.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.157-168
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• 2000

The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H$\infty$, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.

• Journal of Magnetics
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• 제14권2호
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• pp.97-100
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• 2009

A new eddy current testing probe was developed to separate the eddy current signal distortion caused by permeability variation clusters and ordinary defects created in steam generator tubes. Signal processing circuits were inserted into the probe to increase the signal-to-noise ratio and allow digital signal transmission. The new probe could measure and separate the magnetic phases created in the steam generator tubes in the operating environment of a nuclear power plant. Furthermore, the new eddy current testing probe can measure the defects in steam generator tubes as rapidly as a bobbin probe with enhanced testing speed and reliability of defect detection.

• Proceedings of the KIEE Conference
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• 대한전기학회 2003년도 하계학술대회 논문집 D
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• pp.2301-2303
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• 2003

A steam turnine in thermal/nuclear power plant drives generator and maintains it at rated speed using high temperature and high pressure steam energy. After synchronization in parallel with the power system, generator output increases according as the governor, that is the controller, increases steam flow into turbine. By the way, as the steam flow into turbine can not be reduced fast even though the electrical load is lost, the turbine gets into dangerous situation due to the increase of its speed. At this time, the duty of the turbine governor is to limit the speed to its overspeed trip setpoint by stopping the steam flow as soon as possible, the test of which is called load rejection test. It is introduced in this paper for a simulation test of generator load rejection to be implemented on the turbine governor in a 600MW nuclear power plant before its startup.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• 제9권2호
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• pp.125-137
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• 2000

In this paper we propose a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by simulation and experiment for similar steam generator moldel.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 한국공작기계학회 1996년도 추계학술대회 논문
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• pp.107-112
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• 1996

In this paper it is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the preposed vision system, Performance of proposed digital vision system is illustrated by experiment for similar steam generator model.