• The Transactions of the Korean Institute of Electrical Engineers B
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• v.52 no.4
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• pp.147-153
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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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• v.54 no.1
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• pp.152-161
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• 2022

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The "cooldown strategy" is recommended for break flow control, whereas the "RCS make-up strategy" is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.940-947
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• 2022

Meta-heuristic algorithms have found their place in optimization problems. Henry gas solubility optimization (HGSO) is one of the newest population-based algorithms. This algorithm is inspired by Henry's law of physics. To evaluate the performance of a new algorithm, it must be used in various problems. On the other hand, the optimization of the proportional-integral-derivative (PID) gains for load-following of a nuclear power plant (NPP) is a good challenge to assess the performance of HGSO. Accordingly, the power control of a pressurized water reactor (PWR) is targeted, based on the point kinetics model with six groups of delayed-neutron precursors. In any optimization problem based on meta-heuristic algorithms, an efficient objective function is required. Therefore, the integral of the time-weighted square error (ITSE) performance index is utilized as the objective (cost) function of HGSO, which is constrained by a stability criterion in steady-state operations. A Lyapunov approach guarantees this stability. The results show that this method provides superior results compared to an empirically tuned PID controller with the least error. It also achieves good accuracy compared to an established GA-tuned PID controller.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.11a
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• pp.89-103
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• 2002

기존 원자력 발전 기술이 안고 있는 문제점들을 해결함과 동시에 장기적으로 기존의 경ㆍ중수형 원자로를 대체 또는 보완할 수 있을 것으로 기대되는 새로운 원자력에너지시스템으로서 토륨ㆍ우라늄ㆍ플루토늄 혼합 핵주기에 기초한 용융염원자로(Molten Salt Reactor)의 일종인 AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-missioning EXperimental and TEst Reactor) 원자력에너지 복합시스템의 개념설계 연구가 진행중이다.(중략)

• Proceedings of the KWS Conference
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• 2003.11a
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• pp.55-57
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• 2003

The fuel assemblies as the nuclear fuel for the pressurized water reactor(PWR) are loaded in the reactor core throughout the residence time of three to five years. The fuel assembly is manufactured using special welding processes and under strict quality assurance and control systems. In this paper welding parts, welding methods, and welding tests for the integrity of the PWR fuel assemblies are introduced.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.146-149
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• 1994

A nonlinear dynamic optimization has been performed for reactor power control system of CANDU 6 nuclear power plant considering xenon, fuel and moderator temperature feedback effects. Integral-of-Time-multiplied Absolute-Error (ITAE) criterion has been used as a performance index of the system behavior. Optimum controller gain are found by searching algorithm of Sequential Quadratic Programming (SQP). System models are referenced from most recent literatures. Signal flow network construction and optimization have been done by using commercial computer software package.

• JSTS:Journal of Semiconductor Technology and Science
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• v.5 no.3
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• pp.149-158
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• 2005

This paper describes design, fabrication, and application of the silicon based temperature controllable micro reactor. In order to achieve fast temperature variation and low energy consumption, reaction chamber of the micro reactor was thermally isolated by etching the highly conductive silicon around the reaction chamber. Compared with the model not having thermally isolated structure, the thermally isolated micro reactor showed enhanced thermal performances such as fast temperature variation and low energy consumption. The performance enhancements of the micro reactor due to etched holes were verified by thermal experiment and numerical analysis. Regarding to 42 percents reduction of the thermal mass achieved by the etched holes, approximately 4 times faster thermal variation and 5 times smaller energy consumption were acquired. The total size of the fabricated micro reactor was $37{\times}30{\times}1mm^{3}$. Microchannel and reaction chamber were formed on the silicon substrate. The openings of channel and chamber were covered by the glass substrate. The Pt electrodes for heater and sensor are fabricated on the backside of silicon substrate below the reaction chamber. The dimension of channel cross section was $200{\times}100{\mu}m^{2}$. The volume of reaction chamber was $4{\mu}l$. The temperature of the micro reactor was controlled and measured simultaneously with NI DAQ PCI-MIO-16E-l board and LabVIEW program. Finally, the fabricated micro reactor and the temperature control system were applied to the thermal denaturation and the trypsin digestion of protein. BSA(bovine serum albumin) was chosen for the test sample. It was successfully shown that BSA was successfully denatured at $75^{\circ}C$ for 1 min and digested by trypsin at $37^{\circ}C$ for 10 min.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.3
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• pp.338-345
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• 2000

A model testing has been performed to investigate the flow characteristics of a vortex chamber, which plays a role of a flow switch and passively controls the discharge flow rate. This method of passive flow control is a matter of concern in the design of advanced nuclear reactor systems as an alternative to the active flow control to provide emergency water to the reactor core in case of postulated accidents like LOCA (Loss-Of-Coolant Accident). By changing the inflow direction in the vortex chamber and varying the flow resistance inside the chamber, the vortex chamber can control passively the injection flowrate. Fundamental characteristics such as discharge flow rate and pressure drop of the vortex chamber are measured, and its parametric effects on the performance of the vortex chamber are also systematically investigated.

• Transactions of the KSME C: Technology and Education
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• v.3 no.3
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• pp.225-232
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• 2015

The control element drive mechanism (CEDM) is an electro-mechanical device to control reactivity of the nuclear reactor. The conventional CEDM was installed on a nozzle of the reactor vessel closure head as an ex-vessel type. However, there have been demands for an in-vessel CEDM to fundamentally eliminate the rod ejection accident. Conceptual design of the in-vessel CEDM, which was developed based on the existing technology of the ex-vessel CEDM, is introduced in this paper.