• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.858-865
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• 2006

SMART Pilot is a multipurpose small capacity integral type reactor. Main coolant pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of $310^{\circ}C$ and 14.7MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present wort a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and lift-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.827-838
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• 2023

Centrifugal pump is a key part of nuclear power plant systems, and its health status is critical to the safety and reliability of nuclear power plants. Therefore, fault diagnosis is required for centrifugal pump. Traditional fault diagnosis methods have difficulty extracting fault features from nonlinear and non-stationary signals, resulting in low diagnostic accuracy. In this paper, a new fault diagnosis method is proposed based on the improved particle swarm optimization (IPSO) algorithm-based variational modal decomposition (VMD) and relevance vector machine (RVM). Firstly, a simulation test bench for rotor faults is built, in which vibration displacement signals of the rotor are also collected by eddy current sensors. Then, the improved particle swarm algorithm is used to optimize the VMD to achieve adaptive decomposition of vibration displacement signals. Meanwhile, a screening criterion based on the minimum Kullback-Leibler (K-L) divergence value is established to extract the primary intrinsic modal function (IMF) component. Eventually, the factors are obtained from the primary IMF component to form a fault feature vector, and fault patterns are recognized using the RVM model. The results show that the extraction of the fault information and fault diagnosis classification have been improved, and the average accuracy could reach 97.87%.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.504.2-504
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• 2001

• Journal of KSNVE
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• v.9 no.1
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• pp.85-96
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• 1999

For ASME Code pumps in nuclear power plants, inservice test is required to assess the operational readiness in accordance with ASME code and related regulations. The objective of this study therefore, is to develop the technical background of the degradation of pump performances and conditions due to low flow rate operation. In addition. the detection techniques of pump operating conditions are to be developed to enhance the safety and economy of nuclear power plants. A test loop consisted of pump, motor. water tank, flow rate measurements and piping system with flow control devices was established for this study. Two typical pumps, 1-stage volute pump and 3-stage turbine pump, were selected and the test was performed upon two major point of views ; i.e., pump discharge pressure pulsations analysis and pump vibration spectrum analysis. From the test results, it is concluded that (1) the pump vibration affected by the natural frequency of operating pump is significant in the low frequency zone (around 1 Hz) : the vibration amplitude. especially. is an important factor during low flow rate operation. and shall be monitored to ensure that it is within the limit of ASME OM code Part 6, (2) the vibration frequency and pump discharge pressure are affected by vane pass frequency and running speed, (3) the wave phenomena due to the compressiblity of water is anticipated during low flow rate operation. and the pump system shall be designed to prevent it and. finally, (4) the technical background of the degradation of pump performances and conditions due to low flow rate operation is provided.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1310-1323
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• 2023

A reactor coolant pump (RCP) is essential for transporting coolant in the primary loop of pressurized water reactors. In the advanced passive reactor, the absence of a long pipeline between the steam generator and RCP serves as a transition section, resulting in a non-uniform flow field at the pump inlet. Therefore, the characteristics of the pump should be investigated under non-uniform flow to determine its influence on the pump. In this study, the pressure pulsation characteristics were examined in the time and frequency domains, and the sources of low-frequency and high-amplitude signals were analyzed using wavelet coherence analysis and numerical simulation. From computational fluid dynamics (CFD) results, non-uniform inflow has a great effect on the flow structures in the pump's inlet. The pressure pulsation in the pump at the rated flow increased by 78-128.7% under the non-uniform inflow condition in comparison with that observed under the uniform inflow condition. Furthermore, a low-frequency signal with a high amplitude was observed, whose energy increased significantly under non-uniform flow. The wavelet coherence and CFD analysis verified that the source of this signal was the low-frequency pulsating vortex under the steam generator.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.721-722
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• 2005

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2325-2334
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• 2023

To study the lubrication performance of tilting-pad thrust bearing (TPTBs) during start-up in nuclear pump, a hydrodynamic lubrication model of TPTBs was established based on the computational fluid dynamics (CFD) method and the fluid-structure interaction (FSI) technique. Further, a mesh motion algorithm for the transient calculation of thrust bearings was developed based on the user defined function (UDF). The result demonstrated that minimum film thickness increases first and then decreases with the rotational speed under start-up condition. The influence of pad tilt on minimum film thickness is greater than that of collar movement at low speed, and the establishment of dynamic pressure mainly depends on pad tilt and minimum film thickness increases. As the increase of rotational speed, the influence of pad tilt was abated, where the influence of the moving of the collar dominated gradually, and minimum film thickness decreases. For TPTBs, the circumferential angle of the pad is always greater than the radial angle. When the rotational speed is constant, the change rate of radial angle is greater than that of circumferential angle with the increase of loading forces. This study can provide reference for improving bearing wear resistance.

• The KSFM Journal of Fluid Machinery
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• v.2 no.2 s.3
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• pp.13-18
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• 1999

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 we 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.