• 전기의세계
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• 제15권5호
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.

• 대한전기학회논문지:시스템및제어부문D
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• 제54권9호
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• pp.567-569
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• 2005

NARX(Nonlinear AutoRegressive with eXogenous input) neural network was used for prediction of nuclear reactor behavior which was influenced by control rods in short-term period and also by the concentration of xenon and boron in long-term period in load following operations. The developed model was designed to predict reactor power, xenon worth and axial offset with different burnup states when control rods and boron were adjusted in load following operations. Data of the Korea Next Generation PWR were collected by ONED94 code. The test results presented exhibit the capability of the NARX neural network model to capture the long term and short term dynamics of the reactor core and the developed model seems to be utilized as a handy tool for the use of a plant simulation.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 하계학술대회 논문집 B
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• pp.813-815
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• 1998

Engineered Safety Features-Component Control Systems(ESF-CCS) are those I&C systems that control safety equipment used to maintain the integrity of reactor coolant pressure boundary. This paper illustrates distinctive features and improved design concepts of Korea Next Generation Reactor(KNGR) based on the experience obtained through prototyping of ESF-CCS.

• 제어로봇시스템학회논문지
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• 제13권11호
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• pp.1128-1136
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• 2007

This paper presents an application study of a model predictive control based commercial package PCTP to real chemical processes. The first case study concerns a product purity control of a splitter process which distillates styrene from undesired component ethyl-benzene produced from ethyl-benzene dehydrogenation reaction. The second case study is about a temperature control of ethyl-benzene dehydrogenation reactor and an excess oxygen control of the fired heater. Optimum control structure for MPC application is developed for each case study. The application results show a significant improvement in control performance and stability.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.73-78
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• 1996

In this study, we proposed a fuzzy gain scheduler with intelligent learning algorithm for a reactor control. In the proposed algorithm, we used the gradient descent method to learn the rule bases of a fuzzy algorithm. These rule bases are learned toward minimizing an objective function, which is called a performance cost function. The objective of fuzzy gain scheduler with intelligent learning algorithm is the generation of adequate gains, which minimize the error of system. The condition of every plant is generally changed as time gose. That is, the initial gains obtained through the analysis of system are no longer suitable for the changed plant. And we need to set new gains, which minimize the error stemmed from changing the condition of a plant. In this paper, we applied this strategy for reactor control of nuclear power plant (NPP), and the results were compared with those of a simple PI controller, which has fixed gains. As a result, it was shown that the proposed algorithm was superior to the simple PI controller.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.579-585
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• 1996

A CATHENA model for the reactor regulating system is developed and tested independently. A CATHENA plant model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for the Wolsong 2 generating station. This model is intended to provide a trip coverage analysis capability. The CATHENA reactor regulating system model includes the demand power routine. the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model is tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.271-278
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• 2020

The investigation into a full-scale 27 m high, by 6 m wide, thermosyphon loop. The simulation model is based on a one-dimensional axially-symmetrical control volume approach, where the loop is divided into a series of discreet control volumes. The three conservation equations, namely, mass, momentum and energy, were applied to these control volumes and solved with an explicit numerical method. The flow is assumed to be quasi-static, implying that the mass-flow rate changes over time. However, at any instant in time the mass-flow rate is constant around the loop. The boussinesq approximation was invoked, and a reasonable correlation between the experimental and theoretical results was obtained. Experimental results are presented and the flow regimes of the working fluid inside the loop identified. The results indicate that a series of such thermosyphon loops can be used as a cavity cooling system and that the one-dimensional theoretical model can predict the internal temperature and mass-flow rate of the thermosyphon loop.

• 전력전자학회:학술대회논문집
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• 전력전자학회 1998년도 Proceedings ICPE 98 1998 International Conference on Power Electronics
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• pp.276-280
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• 1998

This paper proposes a harmonic a harmonic reduction technique of the parallel-connected twelve-pulse thyristor rectifiers. The proposed system is an improvement over the diode rectifier system with an active interphase reactor [2]. In this scheme, a low KVA (0.15 Po (PU) ) active current source injects a triangular current into an interphase reactor of a twelve-pulse thyristor rectifier along the phase delay angle. The current injection results in near sinusoidal input current with less than 1% THD. Detailed analysis of the proposed scheme along scheme along with design equations is illustrated. Simulation results verify the concept.