• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2010.10a
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• pp.429-430
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• 2010

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.40-46
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• 2016

In the paper, a study on the analysis of fouling effect of the venturi flowmeter is described. In the research flow field solutions about the venturi flowmeter with fouling are obtained and then the effects on fouling states by inserting a ring into the throat of venturi flowmeter are studied. As the result shows, it is found that the inserted ring reduces the fouling effect due to the flow separation occurring at the ring. Consequently, a venturi flowmeter with ring shows smaller pressure loss differences than the original configuration with no ring on fouling state. This research suggests an efficient and economic method of inserting a ring to reduce the pressure loss effects due to fouling.

• The KSFM Journal of Fluid Machinery
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• v.2 no.2 s.3
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• pp.19-26
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• 1999

Thermal output in a nuclear power plant is verified with calorimetric heat balance on the secondary plant. The calorimetry involves the precise measurement of the feedwater flow rate. However, the correct indication of feedwater flow rate obtained by a pressure-difference measurement across a venturi can be affected by instrument errors, fouling or a poorly developed velocity profile. This can result in an inaccurate mass flow rate and consequently an inaccurate estimate of power. The purpose of this study is to develop verification methods with accuracy better than $0.5\%$ for high precision flow measurement to be used for measuring feedwater flow rate. This chemical tracer method is a testing process that uses tracers which can be applied to quantify losses in electrical output due to the incorrect measurements of feedwater flow rate. And this system has good response to the variation of the flow rate. Accuracy of better than 0.5 percent can be expected for feedwater flow measurement, providing that the system can be stabilized during the test. This methodology is applicable to other flow systems well.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.04a
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• pp.648-654
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• 2011

The feed-water piping system constitutes a complex flow impedance network incorporating dynamic transfer characteristics which will amplify some pulsation frequencies. Understanding pressure pulsation waves for the feed-water recirculation piping system with cavitation problem of flow control valve is very important to prevent acoustic resonance. Feed water recirculation piping system is excited by potential sources of the shock pulse waves by cavitation of flow control valve. The pulsation becomes the source of structural vibration at the piping system. If it coincides with the natural frequency of the pipe system, excessive vibration results. High-level vibration due to the pressure pulsation affects the reliability of the plant piping system. This paper discusses the piping vibration due to the effect of shock pulsation by the cavitation of the flow control valves for the recirculation piping of feed-water pump system in combined cycle power plants.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.228-237
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• 1986

This paper provides a comparative study on methodologies for solutions of the inverse problems of certain basic fuzzy relational equations, with which fuzzy set is defined as mapping from sets into complete Brouwerian lattice. Three different algorithms developed so far are discussed and applied to fault diagnosis problem for the main coolant pump of nuclear power plants.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.23 no.12
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• pp.201-209
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• 2009

In power plant operations, sensor values often exhibit erroneous values due to their failures or the intrusions of various noises. However, most of the power plant monitoring and fault diagnosis systems perform their tasks based on the assumptions that the collected sensor values are correct all the times. These assumptions, which are not valid, often lead to serious consequences such as power plant trips. In this paper, we propose a power plant sensor value validation technique that can utilize the relationships existing among the sensor values as the sensor redundancy. The proposed technique is applied to the flow meters installed along boiler feed water systems of a typical tubular type boiler thermal power plant and shows a good potential of future applications.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.19 no.8
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• pp.843-849
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• 2009

The reliable operation of a swing check valve in the main feed water system of a power plant is most essential for successful shout-down process. A failure to close the valve at proper time often leads to the instability of the main feed water system, or even to an emergency stop of the power plant. In reality it is a very difficult task to monitor the behavior of a swing check valve. Furthermore it is impossible to see the motion of the valve. In this work two measurements were carried out simultaneously to determine the precise valve closure time. The dynamic pressure measurements were made at the inlet and outlet regions of the swing check valve. The transient vibration of the valve housing in the direction of water flow was also measured, which enabled the measurement of the transient vibration of the valve housing near valve closure. By comparing the results produced from these measurements the precise valve closure time could be determined. By carrying out order tracking technique using the dynamic pressure signals and pump rpm signal, the complicated dynamic problems inside the main feed water system can be more easily dealt with. This measurement scheme might be implemented in a power plant on a real-time basis without much difficulty. If this could be implemented, valuable information essential for shut-down operations can readily be passed on to the main control room. The feasibility of this implementation was demonstrated by this experimental work.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.310-316
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• 1996

현재 사용중인 Non-LOCA 해석용 인허가 코드들은 특정한 형태의 가압경수로에 맞게 짜여진 것들이어서 모든 형태의 가압경수로에 적용할 수 있는 범용 코드의 개발이 필요한 실정이다. 이를 위하여 한국원자력연구소에서는 웨스팅하우스 및 CE형 발전소에 공히 적용할 수 있는 과도현상 해석 코드인 TASS코드를 개발하고 있다. 이 TASS코드는 실시간보다 빠르게 핵증기계통에 대한 모의계산을 수행하며 대화식의 입출력을 통하여 사용자가 원하는 과도현상을 정확히 모사할 수 있다. 이 TASS코드의 웨스팅하우스형 발전소에 대한 적용타당성은 이미 검증되었으며, 본 논문에서는 CE형 발전소에 대하여 TASS 코드를 적용하여 Non-LOCA 인허가 해석을 하기위한 검증을 위해 주급수관 파단사고 및 주증기관 파단사고에 대하여 RELAP5/MOD3 코드와의 비교계산을 수행하였다.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.