• Nuclear Engineering and Technology
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• 제27권2호
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• pp.155-164
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• 1995

증기발생기 수위조절계통의 무작위추출 비선형 제어계통의 경우로 연구되었다. 무작위 변수로는 시간불연속 계통의 추출시간간격 이 고려되었다. Lyapunov 함수를 구하지 않는 확률론적 Lyapunov 방법이 용되었다. 유도된 안정성 요건은 CANDU 형 원자로인 월성 1호기의 자료를 이용하여 시간 존속 모사로 검증하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1995년도 추계학술대회 논문집 학회본부
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• pp.172-174
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• 1995

About a half of Electric power is generated by nuclear power plants in korea. So, the stable operation of nuclear power plant is very important for suppling the essential national electric power. A S/G(Steam Generator) level control is the most difficult system in PWR(Pressurized Water Reactor) nuclear power plant. Because of the non-linear and the non-nominal response of S/G level control, it Is very difficult to control the level by automatic mode or manual mode. The goal of this study is to establish and verify a advanced control algorithm by analyzing, modelling, stability calculation, controller parameter calculation, simulation for S/G level control system.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.351.1-351
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• 1999

• 대한기계학회논문집B
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• 제24권7호
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• pp.994-1001
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• 2000

The dynamic behavior of a single-pressure heat recovery steam generator and turbine system for the combined cycle power plant is simulated on the basis of one-dimensional unsteady governing equations. A water level control and a turbine power control are also included in the calculation routine. Transient response of the system to the variation of gas turbine exit condition is simulated and effect of the turbine power control on the system response is examined. In addition, the effect of the treatment of inertia terms(fluid inertia and thermal inertia of heat exchanger metal) on the simulated transient response is investigated.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.32-40
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• 1994

중기발생기의 수위조절과 관련하여 최적제어이론을 이용한 디지탈 제어시스템을 설계하였다. 우선 급수변을 일차 지연함수로 취급하여 전체 시스템에 포함시킴으로써 보다 실제에 가까운 시스템이 되게 하였다. LQ 방법을 이용하여 급수변의 작동 및 요구신호량과 급수출력과의 차이를 최소화시킬 수 있는 최적 이득상수를 결정하였으며, 아울러 저출력에서의 급수유량 측정이 불확실함을 고려하여 급수신호에 대한 칼만 관측기를 설계하였다. 그리고 전출력 구간에서 일정한 안정여유도를 유지시킬 수 있는 가변상수 디지탈 제어기를 설계하였다. 이러한 제어시스템은 보다 현실적인 상황을 반영하고 있으며 저출력에서 급수와 반대현상을 보이는 중기발생기의 동특성에도 불구하고 만족할만한 제어특성을 보이고 있다.

• 한국지능시스템학회논문지
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• 제13권3호
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• pp.322-327
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• 2003

가압경수로 원자력 발전소의 증기발생기 수위는 유량의 변동에 상반되는 수축(shrink)과 팽창(swell) 효과 등의 특성을 가지고 있으므로 제어가 어려운 대상으로 알려져 있다. 본 논문에서는 신경망을 이용하여 원자력발전소에서 사용되고 있는 두 개의 PI 제어기 중 부적절한 게인으로 조정된 제어기를 먼저 선택하고, 선택된 제어기의 게인을 퍼지 논리를 적용하여 조정하도록 구성하였다. 게인 조정을 위해 사용되는 기본 정보는 수위, 급수량, 그리고 증기량이다. 이 세 가지의 정보를 바탕으로 신경망을 통해 수위 제어기 또는 급수량 제어기 둘 중 하나의 제어기가 선택한 후 퍼지 자기동조기(self-tuner)를 이용하여 PI 제어기의 게인을 알맞게 조정하게 된다. 퍼지 자기동조기의 규칙은 증기발생기의 상태를 표현하는 입ㆍ출력 데이터의 특성으로부터 추출하였다. 이상의 두 과정을 통해 적절한 제어기를 선택하고, 선택된 제어기의 게인을 알맞게 조정하는 것이 본 논문의 목적이다.

• 연구논문집
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• 통권27호
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• pp.75-86
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• 1997

Combined cycle power plant is a system where a gas turbine or steam turbine is used to produce shaft power to drive a generator for producing electrical power and the steam from the HRSG is expanded in a steam turbine for additional shaft power. Combined cycle plant is a one from of cogeneration. The temperature of the exhaust gases from a gas turbine ranges from $400^\circC$ to $600^\circC$, and can be used effectively in a heat recovery steam generator to produce steam. Combined cycle can be classed as a "topping(gas turbine)" and a "bottoming(steam turbine)" cycle. The first cycle, to which most of the heat is supplied, is called the topping cycle. The wasted heat it produces is then utilized in a second process which operates at a lower temperature level and is therefore referred to as a "bottoming cycle". The combination of gas/steam turbine power plant managed to be accepted widely because, first, each individual system has already proven themselves in power plants with a single cycle, therefore, the development costs are low. Secondly, the air as a working medium is relatively non-problematic and inexpensive and can be used in gas turbines at an elevated temperature level over $1000^\circC$. The steam process uses water, which is likewise inexpensive and widely available, but better suited for the medium and low temperature ranges. It, therefore, is quite reasonable to use the steam process for the bottoming cycle. Only recently gas turbines attained inlet temperature that make it possible to design a highly efficient combined cycle. In the present study, performance analysis of a dual pressure combined-cycle power plant is carried out to investigate the influence of topping cycle to combined cycle performance.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.460-467
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• 2010

In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.

• Nuclear Engineering and Technology
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• 제27권1호
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• pp.8-17
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• 1995

최근 디지탈 기술을 응용하여 원자력 발전소 제어 시스템의 성능을 향상시키려는 많은 노력이 있어 왔고, 차세대 원자로에 구현할 것을 목표로 디지탈 제어 시스템 개발에 관한 장기적인 계획이 수립되어 있다. 디지탈 제어 시스템을 구축하고자 할 때 적정한 샘플링 주기를 정하는 것은 중요한 과정이다. 제어기의 안정성과 성능은 샘플링 주기에 밀접한 관련이 있다 현재 디지탈 제어기의 샘플링 주기를 체계적으로 정하는 전형적인 방법은 없다. 일반적으로 디지탈 제어기의 안정성을 고려해서 연속시간역 제어의 대역폭의 20∼30배의 역수 정도의 샘플링 주기를 흔히 쓴다. 이 논문에서는 안정성 뿐 아니라 시간역에서의 좋은 성능을 보장받을 수 있는 적정한 샘플링 주기를 얻을 수 있는 방법을 제시하였다. 이 방법으로 Irving의 모델을 이용하여 디지탈제어기의 적정 샘플링 주기를 예측하고, 고리 2호기의 마이크로 시뮬레이터와 WDPF 디지탈 제어기를 이용한 증기 발생기 수위제어 모사시스템에서 검증해 본 결과, 고리 2호기 원자력발전소 중기 발생기 수위제어를 위한 디지탈 제어기의 원자로 전 출력영역에 대한 적정 샘플링 주기 가 1초가 되는 것을 알게 되었다.