• Journal of the Korean Society of Propulsion Engineers
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• v.2 no.2
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• pp.74-82
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• 1998

Combined cycle power plant is a system where a gas turbine or a 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. The temperature of the exhaust gases from a gas turbine ranges from $400{\sim}650^{\circ}C$, and can be used effectively in a heat recovery steam generator to produce steam. Combined cycle can be classed as a topping and bottoming cycle. The first cycle, to which most of the heat is supplied, is a Brayton gas turbine cycle. The wasted heat it produces is then utilized in a second process which operates at a lower temperature level is a steam turbine cycle. The combined gas and steam turbine power plant have been widely accepted because, first, each separate system has already proven themselves in power plants as an independent cycle, therefore, the development costs are low. Secondly, using the air as a working medium, the operation is relatively non- problematic and inexpensive and can be used in gas turbines at an elevated temperature level over $1000^{\circ}C$. 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. Recently gas turbine attained inlet temperature that make it possible to design a highly efficient combined cycle. In the present study, performance analysis of a 3 pressured combined cycle power plant is carried out to investigate the influence of topping cycle to combined cycle performance. Present calculation is compared with acceptance performance test data from SeoInchon combined cycle power plant. Present results is expected to shed some light to design and manufacture 150~200MW class heavy duty gas turbine whose conceptual design is already being undertaken.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.457-467
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• 2001

It is well known that the performance of a fuzzy neural network strongly depends on the input features selected for its training. In its applications to sensor signal estimation, there are a large number of input variables related with an output As the number of input variables increases, the training time of fuzzy neural networks required increases exponentially. Thus, it is essential to reduce the number of inputs to a fuzzy neural network and to select the optimum number of mutually independent inputs that are able to clearly define the input-output mapping. In this work, principal component analysis (PCA), genetic algorithms (CA) and probability theory are combined to select new important input features. A proposed feature selection method is applied to the signal estimation of the steam generator water level, the hot-leg flowrate, the pressurizer water level and the pressurizer pressure sensors in pressurized water reactors and compared with other input feature selection methods.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.317-326
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• 1995

The oater level control system of the steam generator in a pressurized water reactor and its control Problems are analysed. In this work a stable control strategy Particularly during low Power operation based on the fuzzy control method is studied. The control strategy employs substitutional information using the bypass valve opening instead of incorrectly measured signal at the low How rate as the fuzzy variable of the flow rate during low power operation, and includes the flexible scale adjusting method for fast response at a large transient. A self-tuning algorithm based on the control performance and the descent method is also suggested for tuning the membership function scale. It gives a practical way to tune the controller under real operation. Simulation was carried out on the Compact Nuclear Simulator set up at Korea Atomic Energy Research Institute and its result showed the good performance of the controller and effectiveness of its tuning.

• Corrosion Science and Technology
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• v.10 no.4
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• pp.151-155
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• 2011

The iron oxide particles could be resulted from the corrosion of the circulating water system of a power plant. Because it may be one of the trouble materials which affect the power generation efficiency due to the deposition on steam generator tube and turbine blade, the continuous observation of its concentration is very important. The laser induced break-down detection (LIBD) technology was applied to monitor continuously the concentration of corrosion products with the detection limit of ppb level. The measurement system consists of a Nd:YAG pulsed laser, a polarizing beam splitter, a flow-type sample cell, an acoustic emission sensor, a high speed data acquisition board, a personal computer, etc.. The performance test results confirmed that this technology can be effective to monitor the corrosion product concentration of the circulating water system of a power plant.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.37-50
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• 1993

To evaluate the sequence of event and the Thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-l/L3-3 experiment. Also, the predictability of the code for the major Thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be performed without core uncovery. It is also found that the plant-specific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance.

• Proceedings of the KIEE Conference
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• 2001.11c
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• pp.11-14
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• 2001

원자력 발전소의 증기 발생기는 증기량과 급수량에 대한 비 최소위상 특성과 비선형성, 그리고 입력 제한 특성을 가지고 있다. 이러한 특성들은 증기 발생기의 효과적인 수위 제어에 어려움을 주고 있다. 본 논문에서는 게인 스케줄링 기법과 변형된 슬라이딩 모드 제어 기법을 이용한 원전 증기 발생기 제어기를 제안한다. 또한 앞먹임 구조를 가진 PI 제어기를 설계하여 저출력 영역에서 제안된 슬라이딩 모드 제어기와 성능을 비교한다. 모의 실험 결과 제안된 슬라이딩 모드 제어기가 최대 수위, 최소 수위, 그리고 안정화 시간 면에서 개선된 성능을 보였다.

• Journal of the Ergonomics Society of Korea
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• v.32 no.1
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• pp.1-7
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• 2013

Objective: This study proposes the application of Ecological Interface Design(EID) method that is effective for situation awareness in digitalized environment. Background: While cognitive interface design method such as Information Rich Display(IRD) is simply focused on existing information for user, EID method helps users' resource to be solved to higher ion task such as diagnostic and problem solving. Method: Using EID method based on Work Domain Analysis (WDA), it was analyzed and designed for Steam Generator(SG) Water Level control process in a digitalized Main Control Room of Nuclear Power Plant. Proposed EID example is evaluated through interviews by expert & operator. Results: The result of expert & operator showed that EID display might give an aid for operator's decision. Conclusion: The results can reduce critical accidental damage that occurred due to cognitive load and so critical human error. Application: This study may be impact on situation awareness study for digitalized interface design.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.53-62
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• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.