• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.921-926
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• 2001

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. In this study, a thorough investigation on the RSPT(Reed Switch Position Transmitter) type control rod position indication system and its actual implementation in the exiting nuclear power plants in Korea was performed first. A design of the control rod position indication system using reed switch for the CEDM on the system-integrated reactor SMART was developed based on the position indicator technology identified through the investigation. The feasibility of the design was evaluated by test of manufactured control rod position indicator using reed switch for SMART.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.79-84
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• 1996

The robust controller for the nuclear reactor power control system is designed. The reactor model is set up by use of the point kinetics equations and the singly lumped energy balance equations. Since the model is different from the actual plant, the controller which makes the system robust is necessary. The perturbation of the actual plant is investigated with respect to several possible sources of uncertainty. Then the overall system is configured into the two port model and the $H_{\infty}$ controller is designed. The loop shaping and the permissible control rod speed are considered as the design constraints. The designed $H_{\infty}$ controller provides the sufficient margins for the robustness, and the system output as well as the control input satisfy their relevant requirements.

• Nuclear Engineering and Technology
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• 제29권4호
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• pp.280-290
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• 1997

The robust controller for the nuclear reactor power control system is designed. The nuclear reactor is modeled by use of the point kinetics equations and the singly lumped energy balance equations, Since the model is not exact, the controller which can make the actual system robust is necessary. The perturbed plant is investigated by employing the uncertainties of the initial power level and the physical properties, and by introducing the delay into the modeled plant The overall system is configured into the two port model and the H$\infty$ controller is designed. In designing the H$\infty$ controller, two factors of the loop shaping and the permissible magnitude of control input are taken into account The designed controller provides the sufficient margins for the robustness, and the transients of the system output power and the control input satisfy their associated requirement.

• 한국산학기술학회논문지
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• 제7권6호
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• pp.1072-1077
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• 2006

본 논문은 AC 라인 리액터와 2개의 4레그 PWM 전압형 컨버터를 사용하여 병렬 및 직렬 능동필터 기능을 가지는 새로운 3상 Line-Interactive UPS 시스템을 제안하고 전압제어 방식을 설명한다. 제안하는 UPS 시스템의 병렬 및 직렬 PWM 전압형 컨버터에서 전압제어의 목적은 정상상태 및 과도상태에서 만족스러운 동작 특성을 보장하는 것이다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 추계학술대회 논문집 전기기기 및 에너지변환시스템부문
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• pp.223-225
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• 2005

This paper presents a novel prototype of a current source resonant inverter using insulated gate bipolar transistors for driving a streamer reactor, streamer generation technology has been recognized as one of the best methods for water treatment, disinfection, industrial wastes utilization, and so on. However, some technological difficulties related to efficient streamer production have been significant problems restricting streamer usage in the industrial plants. Introduced in this paper is a pulse density modulated high frequency inverter for a plasma generate, which is developed with the aim to improve power conversion and control characteristics of the streamer reactor by using advances in power electronic technology. The developed system implements the feedforward control-based pulse density modulation control scheme with pulse width modulation feedback control strategy to compensate temperature and other environmental influences on streamer discharge.

• 전자공학회논문지
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• 제51권9호
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• pp.215-220
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• 2014

본 논문에서는 연구용 원자로의 출력제어기법 설계와 이를 실제 원자로에 적용하여 성능을 검증한 사례를 소개한다. 연구용 원자로의 출력제어를 위해 제안된 설계 원리는 오버슈트(overshoot)의 억제, 출력 증가율의 억제, 그리고 안전해석에 기반한 최대 출력치의 제한이라는 세 가지이며, 이를 만족키 위해 한국원자력연구원 내의 연구용 원자로인 하나로의 설계개념에 기반을 두어 제어 로직의 개념설계, 상세설계, 구현, 시운전을 통해 해외의 원자로에 적용하여 실제 제어 성능을 검증하였다.

• International Journal of Control, Automation, and Systems
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• 제1권2호
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• pp.210-214
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• 2003

This paper presents a residence time distribution (RTD) measurement method using a radioisotope tracer and the estimation method of RTD model parameters to analyze a submerged bio-reactor. The mathematical RTD models have been investigated to represent the flow behavior and the existence of stagnant regions in the reactor. Knowing the parameters of the RTD model is important for understanding the mixing characteristics of a reactor The radioisotope tracer experiment was carried out by injecting a radioisotope tracer as a pulse into the inlet of the reactor and recording the change of its concentration at the outlet of the reactor to obtain the experimental RTD response. The parameter estimation was performed by the Levenberg-Marquardt optimization algorithm. The proposed scheme allowed the parameter estimation of RTD model suggested by Adler-Hovorka with very low deviations. The estimation procedure is shown to lead to accurate estimation of the RTD parameters and to a good agreement between experimental and simulated response.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1088-1099
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• 2021

Time-frequency analysis technique is an effective analysis tool for non-stationary processes. In the field of reactor neutron noise, the time-frequency analysis method has not been thoroughly researched and widely used. This work has studied the time-frequency analysis of the reactor neutron noise experimental signals under bubble disturbance and control rod vibration. First, an experimental platform was established, and it could be employed to reactor neutron noise experiment and data acquisition. Secondly, two types of reactor neutron noise experiments were performed, and valid experimental data was obtained. Finally, time-frequency analysis was conducted on the experimental data, and effective analysis results were obtained in the low-frequency part. Through this work, it can be concluded that the time-frequency analysis technique can effectively investigate the core dynamics behavior and deepen the identification of the unstable core process.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1901-1910
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• 2023

The safety-class (1E) digital control system (DCS) of nuclear power plant characterized structural multiple redundancies, therefore, it is important to quantitatively evaluate the reliability of DCS in different degree of backup loss. In this paper, a reliability evaluation model based on T-S fuzzy fault tree (FT) is proposed for 1E DCS of nuclear power plant, in which the connection relationship between components is described by T-S fuzzy gates. Specifically, an output rejection control system is chosen as an example, based on the T-S fuzzy FT model, the key indicators such as probabilistic importance are calculated, and for a further discussion, the T-S fuzzy FT model is transformed into Bayesian Network(BN) equivalently, and the fault diagnosis based on probabilistic analysis is accomplished. Combined with the analysis of actual objects, the effectiveness of proposed method is proved.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.