• Nuclear Engineering and Technology
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• 제50권6호
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• pp.869-876
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• 2018

The force of a direct current (DC) electromagnetic pump used to transport liquid lithium was analyzed to optimize its geometrical and electrical parameters by numerical simulation. In a heavy-ion accelerator, which is being developed in Korea, a liquid lithium film is utilized for its high charge-stripping efficiency for heavy ions of uranium. A DC electromagnetic pump with a flow rate of $6cm^3/s$ and a developed pressure of 1.5 MPa at a temperature of $200^{\circ}C$ was required to circulate the liquid lithium to form liquid lithium films. The current and magnetic flux densities in the flow gap, where a $Sm_2Co_{17}$ permanent magnet was used to generate a magnetic field, were analyzed for the electromagnetic force distribution generated in the pump. The pressure developed by the Lorentz force on the electromagnetic force was calculated by considering the electromotive force and hydraulic pressure drop in the narrow flow channel. The opposite force at the end part due to the magnetic flux density in the opposite direction depended on the pump geometrical parameters such as the pump duct length and width that defines the rectangular channels in the nonhomogeneous distributions of the current and magnetic fields.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1842-1852
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• 2019

The AP1000 reactor coolant pump is a vertical shielded-mixed flow pump, is the most important coolant power supply and energy exchange equipment in nuclear reactor primary circuit system, whose steadystate and transient performance affect the safety of the whole nuclear island. Moreover, safety demonstration of reactor coolant pump is the most important step to judge whether it can be practiced, among which software simulation is the first step of theoretical verification. This paper mainly introduces the fluid-solid coupling simulation method applied to reactor coolant pump, studying the feasibility of simulation results based on workbench fluid-solid coupling technology. The study found that: for the unsteady calculations of the pure liquid media, the average head of the reactor coolant pump with bidirectional fluid-solid coupling decreases to a certain extent. And the coupling result is closer to the real experimental value. The large stress and deformation of rotor under different flow conditions are mainly distributed on impeller and idler, and the stress concentration mainly occurs at the junction of front cover plate and blade outlet. Among the factors that affect the dynamic stress change of rotor, the pressure load takes a dominant position.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.479-479
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• 2001

• 대한전기학회:학술대회논문집
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• 대한전기학회 2007년도 제38회 하계학술대회
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• pp.614-615
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• 2007

There are roughly a hundred types of cables in power plants. The distribution of circuits in a nuclear plant is comprised of 20% instrument cables, 61% control cables, 13% AC power cables, 1% DC power cables, and 5% communication lines. In the nuclear power plant, medium voltage cables are generally included in the scope of systems reviewed for safety and are included in a plant's maintenance program. Medium voltage cables provide power to many critical components in plants, including feed water pumps, circulating water pumps, and condensate pumps. Among these cables, high temperature sections of cables feeding electrical power to the circulating water pump and the condenser extraction pump were found. The evaluation for these cables is performed to find the maximum allowable current and temperature. The result shows that the load current flowed about 85% of the allowable current ampacity, and the temperature of conductor at full load current did not exceed the limited temperature. Therefore, existing cables for circulating water pump and condenser extraction pump system are going to be used during design life.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.48-55
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• 2024

The safety-related raw water system's strong operational condition supports the radiation defense and biological shield of nuclear plant containment structures. Gaps and failures in maintaining proper working condition of main equipment like pump were among the most common causes of unavailability of safety related raw water systems. We integrated the advanced data analytics tools to evaluate the maintenance records of water systems and gave special consideration to deficiencies related to pump. We utilized maintenance data over a three-and-a-half-year period to produce metrics like MTBF, MTTF, MTTR, and failure rate. The visual analytic platform using tableau identified the efficacy of maintenance & deficiency in the safety raw water systems. When the number of water quality violation was compared to the other O&M deficiencies, it was discovered that water quality violations account for roughly 15% of the system's deficiencies. The pumps were substantial contributors to the deficit. Pump availability was predicted and optimized with real time data using response surface method. The prediction model was significant with r-squared value of 0.98. This prediction model can be used to predict forth coming pump failures in nuclear plant.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.40-46
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• 1998

A set of experiments using a 1/10 scale model pump which was manufactured to simulate performance of reactor coolant pump(RCP) of Y.G.N # 3 and 4, was executed in single phase(at atmospheric pressure and room temperature) and near-saturation(300 ~ 600kPa). The pump characteristics in single phase flow was similar to the characteristics of the RCP. The pump characteristic curves at nearly saturated state were correlated in terms of flow coefficient and head coefficient for subcooled temperature using the cavitation number defined as (equation omitted), which can be predicted the cavitation possibility. The pump behavior around the saturated temperature almost consists with single phase behavior until the cavitation occurs(When cavitation occurs. When the flow coefficient is about 0.12), the pump head rapidly degrades. In this situation, subcooled temperature is about 1.8~8$^{\circ}C$ and cavitation number of model pump is 1.0 ~ 1.7.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.353-363
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• 2023

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 1998년도 유체기계 연구개발 발표회 논문집
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• pp.7-10
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• 1998

The pump safety related to the functions in nuclear power plants must be designed to meet load conditions considering seismic requirements. In order to satisfy both structural integrity and operability of these pumps, the initial step in the seismic qualification is to establish the resonant frequencies of the structure. Applications are made to the design of the vertical and horizontal type pump. Computational results are analyzed with respect to the dynamic characteristics and are compared to the expected design requirements.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• 한국초전도ㆍ저온공학회논문지
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• 제3권2호
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• pp.15-20
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• 2001

This paper describes a model flux pump experiment recently performed at the MIT Francis Bitter Magnet Laboratory. The results of the model flux pump will be used in the development of a prototype flux pump that will be couple to a high-temperature superconductor (HTS) insert coil of a high-field NMR (Nuclear Magnetic Resonance) magnet, Such an HTS insert is unlikely to operate in persistent model because of the conductors low index(n) The flux pump can compensate fro field decay in the HTS insert coil and make the insert operate effectively in persistent mode . The flux pump, comprised essentially of a transformer an two switches. all made of superconductor, transfers into the insert coil a fraction of a magnetic energy that is first introduced in the secondary circuit of the transformer by a current supplied to the primary circuit. A model flux pump has been designed. fabricated, and operated to demonstrate that a flux pump can indeed supply a small metered current into a load superconducting magnet. A current increment in the range of microamperes has been measured in the magnet after each pumping action. The superconducting model flux pump is made of Nb$_3$ Sn tape, The pump is placed in a gaseous environment above the liquid helium level to keep its heat dissipation from directly discharged in the liquid: the effluent helium vapor maintains the thermal stability of the flux pump.