• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.23 no.1
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• pp.42-47
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• 2010

The power burden of high-$T_c$ superconducting (HTSC) elements comprising superconducting fault current limiter (SFCL) using magnetic coupling of shunt reactors was analyzed. The magnetically coupled shunt reactors play a role in distributing the even power burden between HTSC elements comprising the SFCL, which contributes to the effective current limiting and recovery characteristics of the SFCL. It was confirmed through the comparative analysis on the SFCLs with both the magnetically coupled and the magnetically uncoupled shunt reactors that the magnetically coupled shunt reactors could improve the SFCL's performance by equalizing the power burden of HTSC elements.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.877-887
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• 2022

A new reactor concept is described that directly couples a supercritical CO₂ (sCO₂) power cycle with a CO₂-cooled, heavy water moderated pressure tube core. This configuration attains the simplification and economic potential of past direct-cycle sCO₂ concepts, while also providing safety and power density benefits by using the moderator as a heat sink for decay heat removal. A 200 MWe design is described that heavily leverages existing commercial nuclear technologies, including reactor and moderator systems from Canadian CANDU reactors and fuels and materials from UK Advanced Gas-cooled Reactors (AGRs). Descriptions are provided of the power cycle, nuclear island systems, reactor core, and safety systems, and the results of safety analyses are shown illustrating the ability of the design to withstand large-break loss of coolant accidents. The resulting design attains high efficiency while employing considerably fewer systems than current light water reactors and advanced reactor technologies, illustrating its economic promise. Prospects for the design are discussed, including the ability to demonstrate its technologies in a small (~20 MWe) initial system, and avenues for further improvement of the design using advanced technologies.

• Nuclear Engineering and Technology
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• v.50 no.6
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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.

• Journal of Electrical Engineering and Technology
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• v.9 no.4
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• pp.1369-1374
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• 2014

Most accidents of medium-voltage cables installed in nuclear power plants result from the initial defect of internal insulators or the initial failure due to poor construction. However, as the service years of plants increase, the possibility of cable accidents is also rapidly increases. This is primarily caused by electric, mechanical, thermal, and radiation stresses. Recently, much attention is paid to the study of cable diagnoses. To date, partial discharge and Tan${\delta}$ measurements are known as reliable methods to diagnose the aging of medium-voltage cables. High frequency partial discharge measurement techniques have been widely used to diagnose cables in transmission and distribution systems. However, the on-line high frequency partial discharge technique has not been used in the nuclear power plants because of the plant shutdown risk, degraded measurement sensitivity, and application problems. In this paper, the partial discharge measurement with a portable device was tried to evaluate the integrity of the 4.16kV and 13.8kV cable lines. The test results show that the high detection sensitivity can be achieved by the high frequency partial discharge technique. The present technique is highly attractive to diagnose medium voltage cables in nuclear power plants.

• Journal of the Korean Vacuum Society
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• v.10 no.4
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• pp.440-448
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• 2001

The operation properties of DBD plasma reactors were observed by using 20 kV square pulse at the cylindrical and planar type of reactors in the condition of air pressure. The optimum operation frequency $f_0$ which optimizes the efficiency of operation was found as such $f_0\proptoexp(-C)$ when the current-voltage curve and charge-voltage curve were observed. Using these properties the dissipated power was evaluated. The dissipated power at the optimum frequency of operation was varied as the value of capacitance which is dependent on the structure and the dielectric material of the reactor, and had the maximum value at the specific value of capacitance. With these value of capacitance, DBD reactors which has a high level of efficiency can be formed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.39-48
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• 2024

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1015-1021
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• 2016

Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.433-446
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• 2005

The development of accelerator science and technology has been accommodating ever increasing demand from scientific community of the beam energy and intensity of proton beams. The use of high-powered proton beams has extended from the traditional application of nuclear and high-energy physics to other applications, including spallation neutron source replacing nuclear reactor, nuclear actinide transmutation, energy amplification reactors. This article attempts to review development of proton accelerator, both linear and circular, and issues related to the proton beam energy, intensity as well as its output power. For related accelerator physics and technical review, one should refer to the recent article in the Reviews of Modem Physics [1]

• Journal of Radiation Industry
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• v.18 no.1
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• pp.23-33
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• 2024

This study was conducted to analyze the characteristics of three high-exposure tasks performed by radiation workers in Korean pressurized water reactors (PWRs) and to identify factors that reduce their exposure during work. Three high-exposure tasks were selected based on a previous study. In this previous study, nozzle dam installation and removal, eddy current testing, and manway opening and closing were determined as high-exposure tasks through normalization (radiation dose per unit time). Based on the analysis of the characteristics of the high-exposure tasks in this study, the high-exposure tasks were steam generator-related tasks performed inside and outside the water chamber. This study analyzed the reduction factors for high-exposure tasks and suggested improvements in terms of time, distance, and shielding. The use of the characteristics of high-exposure tasks and their dose reduction factors enables Korean PWRs to optimize radiation protection for workers who receive relatively high doses.