• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.26 no.8
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• pp.65-70
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• 2012

Induction motor is the most widely used to obtain the driving force in the industrial site. Induction motor generates a high current at startup. Most of starting currents are often more than five times of rated current. This high starting current can cause problems such as the voltage drop in the system. In order to solve these problems, if the motor capacity is large, generally we use reactor starting method rather than direct on line starting method. When a high startup current passes through reactor, reactor can serve as a nonlinear elements. In this study, we analyzed that the current, torque and power of the induction motor are different from the change of linear and nonlinear components of the reactor magnetic field.

• Proceedings of the KSME Conference
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• 2000.11b
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2002.02a
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• pp.326-329
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• 2002

The dc reactor type superconducting fault current limiter is composed of a power converter, magnetic core reactors and a do reactor that is a superconducting coil. When a fault occurs, the dc reactor maintains the stability of system by limiting its fault current. In this study, we focus on the design of the dc reactor using FEM(Finite Element Method). In order to design it, various elements should be considered such as magnetic field intensity, Lorentz's force, its inductance and so forth. Firstly, we forecast the values of those elements from the simulation of FEM and then measured with a copper wire magnet. Finally, verify the reliability of this FEM method by comparing with two results.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.921-928
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• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.221-230
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• 2007

Wavelet theory was applied to detect a singularity in a reactor power signal. Compared to Fourier transform, wavelet transform has localization properties in space and frequency. Therefore, using wavelet transform after de-noising, singular points can easily be found. To test this theory, reactor power signals were generated using the HANARO(a Korean multi-purpose research reactor) dynamics model consisting of 39 nonlinear differential equations contaminated with Gaussian noise. Wavelet transform decomposition and de-noising procedures were applied to these signals. It was possible to detect singular events such as a sudden reactivity change and abrupt intrinsic property changes. Thus, this method could be profitably utilized in a real-time system for automatic event recognition(e.g., reactor condition monitoring).

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.9
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• pp.215-220
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• 2014

Study and application result of power controller to research reactor is presented. Considering safety-oriented design concept and other control environment, we developed a simple closed-loop controller that provides limiting function of power-change-rate as well as low-overshoot and fine tracking performance. The design result has been well-proven via simulation and actual application to a research reactor.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1696-1710
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• 2017

An improved Reactor Regulating System (RRS) logic architecture, which is combined with genetic algorithm (GA), is implemented in this work. It is devised to provide an optimal solution to the current RRS. The current system works desirably and has contributed to safe and stable nuclear power plant operation. However, during the ascent and descent section of the reactor power, the RRS output reveals a relatively high steady-state error, and the output also carries a considerable level of overshoot. In an attempt to consolidate conservatism and minimize the error, this work proposes to apply GA to RRS and suggests reconfiguring the system. Prior to the use of GA, reverse engineering is implemented to build a Simulink-based RRS model. Reengineering is followed to produce a newly configured RRS to generate an output that has a reduced steady-state error and diminished overshoot level. A full-scope APR1400 simulator is used to examine the dynamic behaviors of RRS and to build the RRS Simulink model.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.157-164
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• 2015

Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.858-865
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• 2006

SMART Pilot is a multipurpose small capacity integral type reactor. Main coolant pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of $310^{\circ}C$ and 14.7MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present wort a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and lift-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP.