• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1873-1874
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• 2006

증기터빈 발전소에서 터빈제어밸브는 운전중 적절히 개도를 조절하여 보일러에서 공급되는 유량을 조절하여 터빈속도 및 발전기의 전기적 출력을 제어할 수 있도록 해준다. 이러한 터빈제어밸브는 4개로 구성되어 있으며, 운전중 최적의 발전효율을 얻기 위해 저부하에서는 4개의 제어밸브가 동일하게 조절되다가 밸브 전환운전을 완료한 후 4개의 밸브가 주어진 특성에 따라 각기 조절되어야 할 필요성이 있다. 대상 발전소에서 시행하고 있는 운전절차중 하나인 밸브전환 운전은 발전기 출력의 변화가 수반되며 이를 최소화하기 위해 고압터빈측의 1단 압력을 궤환하여 보상하는 알고리듬을 사용하고 있다. 사용중에 발전기 출력변화가 발생되어 이를 개선하기 위해 적용되는 알고리듬을 보완하여 실제 발전소 현장에 적용한 결과를 소개하고자 한다.

• Nuclear Engineering and Technology
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• v.12 no.1
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• pp.9-18
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• 1980

Reflood experiments under atmospheric pressure have been conducted with a single heated tube to investigate basically the rewetting phenomena following a LOCA. Experimental conditions are 180cm length of test tube, wall temperature range of 300-80$0^{\circ}C$, coolant flooding rate of 5-30cm/sec. and subcooling of 35-85$^{\circ}C$. Experiments show that the rewetting velocity is dependent on the initial wall temperature of test tube, coolant flow rate and coolant subcooling. It is required to develop the proper method to evaluate the rewetting temperature and the heat transfer coefficient.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.1669-1670
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• 2008

발전소를 포함한 플랜트 공정 제어로직 설계에서 제어대상인 공정의 특성을 파악하는 일이 중요한 출발점이라고 할 수 있다. 이 논문에는 국내 가동 중인 500MW 석탄화력발전소의 보일러 제어시스템을 교체할 목적으로 우선 현재의 제어로직을 분석한 내용을 기술한 것이다. 대상 발전소의 보일러는 관류형 초임계압 형식으로서 1990년대 초부터 건설되기 시작한 국내 표준모델 중 하나이다. 표준관류보일러의 일반특성과 제어원리에 대하여 고찰한 내용을 먼저 기술하였으며, 효과적인 제어목표를 달성하기 위해 보일러 특성을 제어로직에 반영한 증기압력제어, 급수유량제어 제어알고리듬을 차례로 기술하였다. 여기에 근거 자료 또는 참고자료로 사용된 그래프 등은 고찰대상인 실제 발전소에서 수집하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.751-756
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• 1998

고리 1호기 원전 수명 연장을 위한 가압열충격(Pressurized Thermal Shock : PTS) 해석은 확률론적 안전성 평가 방법에 따라 수행된다. 본 연구는 가압열충격 상세 해석 연구의 일환으로 가압열충격 해석을 위한 계통해석시 사용되는 최적 평가(Best Estimate) 방법과 기존의 PCT(Peak Cladding Temperature) 관점의 해석에 사용되는 결정론적 안전성 평가 방법간의 해석 방법론 차이에 의한 열수력 거동의 상이점을 평가하기 위함이다. 이를 위해 1998년 설치 예정인 고리 1호기 교체 증기발생기(Replacement Steam Generator ; RSG) 안전성 분석 보고서$^{[1]}$ 의 주증기관 파단사고 해석 결과와 동일한 파단 크기 및 운전 출력에 대해 최적 평가 방법론에 따라 해석된 본 연구의 해석 결과를 비교, 평가하였다. 해석 결과 전출력 소형 주증기관 파단 사고에서는 터빈 유량 모델링 및 반응도 계수, 고온 영출력 대형 파단 사고에서는 가압기 모델, 반응도 계수 및 정지여유도가 해석 방법론에 따른 열수력 거동의 차이에 영향이 큰 것으로 평가되었다

• Journal of the Korea Academia-Industrial cooperation Society
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• v.16 no.11
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• pp.7736-7744
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• 2015

Considering the significant waste of industrial energy, effective use of low temperature waste heat is extremely important. In this study, a heat pump cycle with double effect and double stage was realized, which escalates the hot water temperature from $50^{\circ}C$ to $70^{\circ}C$ using $160^{\circ}C$ high temperature heat source and $17^{\circ}C$ low temperature heat source. The steam generated in the first generator condenses in the first condenser generating steam in the second generator. The steam condenses in the second condenser and is provided to the second evaporator. Part of the water out of the second evaporator is supplied to the first evaporator, which evaporates using low temperature waste heat. The evaporated steam enters the first absorber and the second evaporator. The steam out of the second evaporator is absorbed into the solution at the second absorber. The hot water temperature is raised in the second condenser and in the second absorber. Proper flow rates and UA values, which satisfied temperature lift $20^{\circ}C$ and COP 1.6, were deduced through trior and error. The COP increases as the temperature of the high temperature water increases, hot water temperature decreases and flow rate increases, waste water temperature and flow rate increases, solution circulation rate decreases. On the other hand, the temperature rise of the hot water increases as the temperature of the high temperature water increases, hot water temperature increases and flow rate decreases, waste water temperature and flow rate increases, solution circulation rate increases. In addition, the COP and hot water temperature rise increase as UAs of the heat exchangers increase.

• Korean Journal of Food Science and Technology
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• v.22 no.4
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• pp.479-485
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• 1990

For the automation of a evaporation process, computer based evaporation system was built and applied to acquisition of the process variables with an centrifugal thin film evaporator(Centri-Therm, CT-1B). Controls of the process conditions were performed by computer system for pressure, feeding rate, steam, evaporation temperature and flow rate of cooling water. The data acquisitions were also performed by computer system for the changes in the concentration and temperature readings for steam, evaporation and cooling water at the both inlet and outlet. The control and the acquisition variables were collected through the interface device and analyzed by programs using the PASCAL language. To control the feeding rate during the concentration process, inverter was used. The cooling water for the vapor condensation was controlled by the valve controller and should be supplied with the flow rate of 125 kg/h. The maximum vapor condensation rate was 41.7kg/h at the feeding rate of 125 kg/h.

• 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.

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.9-16
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• 1986

Simulation of the system thermal-hydraulic parameters was carried out following the KNUl (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018.

• Membrane Journal
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• v.22 no.3
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• pp.178-190
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• 2012

Flat sheet membranes were prepared with polysulfone (PSF) by an immersion precipitation phase inversion method. Membranes were prepared with PSF/N-methylpyrrolidone (NMP)/polyvinylpyrrolidone (PVP)/phosphoric acid casting solution and water coagulant. By using the successive process of the vapor-induced phase inversion (VIPS) followed by the nonsolvent-induced phase inversion (NIPS), the effect of phosphoric acid addition to casting solution on morphology and permeability of membrane was studied. The mean pore size, the porosity, and the water flux of membranes were increased by the addition of small amount of phosphoric acid. Furthermore, the morphology of the prepared membranes were changed from a dense sponge-like structure to highly enhanced asymmetric structure. PSF/NMP/PVP/phosphoric acid/2-butoxyethanol (BE) casting solution were prepared and cast the successive VIPS-NIPS process with same experimental condition. Due to the addition of BE to casting solution, the mean pore size and almost 0.1 ${\mu}m$ and the water flux increased about 10 to 12 $L/cm^2{\cdot}min{\cdot}bar$.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.483-490
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• 1995

The objective of the PKL III A 4.4 experiment is to examine that the plant could be controlled by manually operative actions "after Steam Generator Tube Rupture under Offsite Power Available without Safety Injection". In order to verify the limitation and ability of the system code NLOOP in the expeiment simulation, the behaviors of the PKL III facility obtained in the experiment are compared with the results of NLOOP code. NLOOP code, which is originally developed to simulate the transients of the Westinghouse type PWRs by KAERI/SIEMENS, modified properly to simulate the PKL III facility. Particular attention is given to the RCS mass How rate of the natural circulation in loops and the termination behavior of the natural circulation in the isolated loop. The comparisons between the experimental and calculational results show the simulation ability and problems of the code. the code.