• Journal of the Korea Academia-Industrial cooperation Society
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• v.5 no.3
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• pp.227-231
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• 2004

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the steam generator or the pump whose type is the axial flow. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of the pump power of the axial flow pump has been introduced in this study. Up to now, we did not found out a precedent which the pump power is used for the flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the flow-rate calculation method by the measurement of the pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs. So, it has been concluded that it is possible to calculate the flow-rate by the measurement of the pump motor inputs.

• Journal of the Korean Vacuum Society
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• v.14 no.2
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• pp.69-77
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• 2005

Pumping performance of large cryosorption pumps of different types installed on the 60 $m^3$ test stand for developing and testing ion sources and beam line components of the NBI system was investigated. Hydrogen adsorption and desorption characteristics of the cryosorption panels were analyzed using the temporal change of the hydrogen spectrum obtained with short introduction of the hydrogen gas as cooling the panel, and simulations on the mutual influence between related parameters were also carried out.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.657-658
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• 2004

The reactor protection system (RPS) of HANARO is a safety class system. The reactor is tripped by dropping four shut off rods (SOR). The SOR system consists of a SOR, hydraulic pump, hydraulic cylinder, solenoid valves and a power supply unit which has the AC coil contactor as a switching component. The hydraulic pump lifts up the SOR. The SOR drops by loss of the hydraulic pressure in the hydraulic circuit at the occurrence of voltage sags or interruptions. From this experiment, we knew that the magnitude of the voltage sag which impacts on this system is 70V, 500msec. The reactor regulation system (RRS) of HANARO has four CARs which are connected to the driver through a magnetic clutch. The CAR drops by loss of electromagnetic force of the magnetic clutch when the deeper voltage sags to lower than 10V, 500msec.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.44-51
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• 2023

The RCP is one of the main components in nuclear power plants and plays an important role in circulating coolant to the RCS system. Currently, nuclear plants are monitored using various monitoring systems. However, since they operate independently according to their functional purpose, it is not able to analyze vibration and operation/performance information comprehensively, and thus failure diagnosis accuracy is limited. In addition, these systems do not provide some important information (such as fault type, parts and cause) necessary for emergency actions, but provide only alarm information. To improve these technical problems, this study proposes a diagnosis technique (M/L, Rule-based model, Data-driven model, Narrow band model) and methodology for comprehensive analysis.

• The KSFM Journal of Fluid Machinery
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• v.3 no.2 s.7
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• pp.24-30
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• 2000

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seen that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stress predictions.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.636-642
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• 2005

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.11a
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• pp.264-269
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• 2008

This paper presents the summary of control of abnormal vibration found in the COP motors of a nuclear power plant. All six identical units of COP pump-motor assemblies showed unstable vibration pattern of which one or two showed higher vibration enough to exceed the allowable level from the installation stage. Many trials of test, measurement, overhaul and replacement had been repeated to investigate and solve the problem but only to reach unsatisfactory settlement. Recently several times of site tests are made and followed by significant diagnostic actions in which the authors group participated. It was found that the coupled shafting system of motor and pump is in close resonance with the $1^{st}$ shaft rotating speed. Redesign of topside motor bearing clearance is made to increase bearing stiffness and hence to avoid the resonance which consequently led to reduce the troubled vibration to allowable and stable status.

• Journal of the Korean Society of Systems Engineering
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• v.9 no.1
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• pp.65-84
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• 2013

This paper introduces a systems engineering approach to reliability centered maintenance to address some of the weaknesses. Reliability centered maintenance is a systematic, disciplined process that produces an efficient equipment management strategy to reduce the probability of failure [1]. The study identifies the need for RCM, requirements analysis, design for RCM implementation. Value modeling is used to evaluate the value measures of RCM. The system boundary for the study has been selected as containment spray pump and its motor drive. Failure Mode and Criticality Effects analysis is applied to evaluate the failure modes while the logic tree diagram used to determine the optimum maintenance strategy. It is concluded that condition based maintenance tasks should be enhanced to reduce component degradation and thus improve reliability and availability of the component. It is recommended to apply time directed tasks to age related failures and failure finding tasks to hidden failures.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.444-447
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor (PWR) or a heavy water power reactor (CANDU). There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The pump can not continuously suck a fluid and not pressurize the fluid during a cold function test. To verify the flow characteristics of the MCWS, a flow net work analysis has been conducted. When the higher elevation pipelines wholly filled with coolant, it was confirmed through the analysis results that the pump pressurized the coolant normally. And the analysis results described the system characteristics with operation temperature and pressure variation satisfactorily.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2183-2195
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• 2020

A mathematical model for the flowrate and rotation speed of RCP during idling was established. The numerical calculation method and dimensionless method were used to analyze the flow, head, torque and pressure and speed changes under idle conditions. Regularity, using the Q criterion vortex identification judgment method combined with surface flow spectrum morphology analysis to diagnose the vortex dynamic characteristics on RCP blade. On impeller blade, there is two oscillations in the pressure ratio on pressure surface in blade outlet region. The velocity on the suction surface is two times more oscillating than the inlet of blade, and there is an intersection with the velocity ratio curve on pressure surface. On blade of guide vane, the pressure ratio increases along the inlet to outlet direction, and the speed ratio decreases with the increase of idle time. There is a vortex that rotates counterclockwise on the suction surface, and the streamline on the suction surface of blade is subjected to the entrainment and blocking action of the vortex creates a large reverse flow in the main flow region. There are two vortices at the outlet of guide vane suction side and the vortices are in opposite directions.