• 대한전기학회:학술대회논문집
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• 대한전기학회 2004년도 학술대회 논문집 정보 및 제어부문
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• pp.641-643
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• 2004

Liquid Zone Control System controls the power of heavy water reactor. Changing the level of each zone compartment regulates one local zone power of 14 zone powers, iud the level is limited less than 90% by the control algorithm to prevent the flood. In recent years, the level and the power was controlled oscillatory in the upper zones. To find out the condition of cycling, the zone control system was modelled with the linear difference equations and identified using parameter estimation. The pole-zero plot showed that the major pole was near the stability boundary, and the system had oscillatory characteristics in nature.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2002년도 ICCAS
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• pp.69.5-69
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• 2002

1. Introduction 2. System Description 3. Hybrid Modeling of SM Reactor 4. Simulation Results 5. Optimization of Operating Condition 6. Conculsion

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2002년도 ITC-CSCC -2
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• pp.1208-1211
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• 2002

The robust controller for the nuclear reactor power control system is designed. The reactor model is described in the form of transfer function and the bound of each coefficient is determined to set up the linear interval system. By the Kharitonov and the edge theorem, a frequency based design template is made and applied to the determination of the controller. The controller designed by this method is simpler than that obtained by the H(equation omitted). Although the controller is designed with the basis of high power, it could be used even at low power.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1997년도 추계학술대회 논문집 학회본부
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• pp.484-486
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• 1997

In this paper, we compose of the utility phovoltaic system with the stepdown chopper and the current source inverter to reduce pulsation of dc current and dc reactor. We study that control several ways(twice frequency of utility voltage square pulse and without the chopper) of chopper part. Therefore if voltage across the dc reactor equal to zero, it is decreased. And we control modulation factor of the chopper to operate at maximum power point around of solar cell.

• Corrosion Science and Technology
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• 제5권3호
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• pp.94-98
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• 2006

The next generation nuclear power reactor will use Pb-Bi as the cooling material. The steel structure materials such as HT-9 used in the reactor suffer from corrosion when they are exposed to high temperature Pb-Bi. Therefore corrosion should be prevented to use Pb-Bi as the coolant material without any safety problem. One method is to control the oxygen content in Pb-Bi. An appropriate amount of oxygen in Pb-Bi can produce a thin oxide layer on steel, and this layer protects the steel from corrosion attack. Since the required oxygen content in Pb-Bi is in the range of $10^{-5}$ to $10^{-7}$ wt%, this small oxygen content can be controlled by flowing a mixture of hydrogen gas and water vapor. The stagnant corrosion test of HT-9 samples was performed by controlling the oxygen content up to 2,000 hours. The corrosion behavior of HT-9 was analyzed at the temperatures of $500^{\circ}C$ and $650^{\circ}C$ with a reduced condition and a oxygen content of $10^{-6}$ wt%.

• 대한전기학회논문지:전력기술부문A
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• 제52권4호
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• pp.147-147
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• 2003

This paper describes the finite element analysis(FEA) for the design of electromagnet for Control Element Drive Mechanism(CEDM) in System-integrated Modular Advanced Reactor(SMART) and compared with the lifting power characteristics of prototype electromagnet. A thermal analysis was performed for the electromagnet. A model for the thermal analysis of the electromagnet was developed and theoretical bases for the model were established. It is important that the temperature of the electromagnet windings be maintained within the allowable limit of the insulation. since the electromagnet of CEDM is always supplied with current during the reactor operation. So the thermal analysis of the winding insulation which is composed of polyimide and air were performed by finite element method. As a result, it is shown that the characteristics of prototype electromagnet have a good agreement with the results of FEA. The thermal properties obtained here will be used as input for the optimization analysis of the electromagnet.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.877-885
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• 2018

Various controllers such as proportional-integral-derivative (PID) controllers have been designed and optimized for load-following issues in nuclear reactors. To achieve high performance, gain tuning is of great importance in PID controllers. In this work, gains of a PID controller are optimized for power-level control of a typical pressurized water reactor using particle swarm optimization (PSO) algorithm. The point kinetic is used as a reactor power model. In PSO, the objective (cost) function defined by decision variables including overshoot, settling time, and stabilization time (stability condition) must be minimized (optimized). Stability condition is guaranteed by Lyapunov synthesis. The simulation results demonstrated good stability and high performance of the closed-loop PSO-PID controller to response power demand.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• Biotechnology and Bioprocess Engineering:BBE
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• 제5권6호
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• pp.456-459
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• 2000

A pilot scale experiment was performed for a year to develop a two-phase anaerobic process for piggery wastewater treatment (COD: 6,000mg/L, BOD: 4,000mg/L, SS: 500mg/L, pH 8.4, alkalinity 6,000mg/L). The acidogenic reactor had a total volume of 3㎥, and the methanogenic reactor, an anaerobic up-flow sludge filter, combining a filter and a sludge bed, was also of total volume 3㎥(1.5㎥ of upper packing material). Temperatures of the acidogenic and methanogenic reactors kept at 20$^{\circ}C$ and 35$^{\circ}C$, respectively. When the pH of the acidogenic reactor was controlled at 6.0-7.0 with HCl, the COD removal efficiency increased from 50 to 80% over a period of six months, and as a result, the COD of the final effluent fell in the range of 1,000-1,500 mg/L. BOD removal efficiency over the same period was above 90%, and 300 to 400 mg/L was maintained in the final effluent. The average SS in the final effluent was 270 mg/L. The methane production was 0.32㎥ CH$_4$/kg COD(sub)removed and methane content of the methanogenic reactor was high value at 80-90%. When the pH of the acidogenic reactor was not controlled over the final two months, the pH reached 8.2 and acid conversion decreased compared with that of pH controlled, while COD removal was similar to the pH controlled operation. Without pH control, the methane content in the gas from methanogenic reactor improved to 90%, compared to 80% with pH control.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1110-1119
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• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.