• 한국IT서비스학회지
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• 제20권3호
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• pp.1-12
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• 2021

Reactor Coolant Pump (RCP) is core part of nuclear power plant to provide the forced circulation of reactor coolant for the removal of core heat. Properly monitoring vibration of RCP is a key activity of a successful predictive maintenance and can lead to a decrease in failure, optimization of machine performance, and a reduction of repair and maintenance costs. Here, we developed real-time RCP Vibration Analysis System (VAS) that web based platform using NoSQL DB (Mongo DB) to handle vibration data of RCP. In this paper, we explain how to implement digital signal process of vibration data from time domain to frequency domain using Fast Fourier transform and how to design NoSQL DB structure, how to implement web service using Java spring framework, JavaScript, High-Chart. We have implement various plot according to standard of the American Society of Mechanical Engineers (ASME) and it can show on web browser based on HTML 5. This data analysis platform shows a upgraded method to real-time analyze vibration data and easily uses without specialist. Furthermore to get better precision we have plan apply to additional machine learning technology.

• 방사성폐기물학회지
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• 제16권3호
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• pp.389-396
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• 2018

국내 가동원전 중 2-루프 가압경수로인 고리1호기는 약 40년 운전한 후, 2017년 6월 18일 영구정지되었다. 영구정지된 고리 1호기는 주요 해체작업을 수행하기전에 계통내 선량률을 저감시켜 작업자피폭을 최소화하기 위한 계통제염을 수행할 예정이다. 일반적으로, 계통제염 범위는 원자로압력용기, 가압기, 증기발생기, 화학 및 체적제어계통, 잔열제거계통 및 원자로 냉각재계통 주요배관을 포함한다. 이러한 계통 및 기기 등을 효율적으로 제염하기 위해서는 제염과정에서 원자로냉각재계통내 유동특성을 평가할 필요가 있다. 계통제염을 위해 순환유량을 제공하는 방법은 다양하나, 본 논문에서는 잔열제거펌프 운전에 따른 고리1호기 원자로냉각재계통내 유동특성을 평가하였다. 잔열제거펌프를 이용한 계통제염은 원자로냉각재 내 유량의 불균형을 초래하여 계통내 기기 및 배관 등에 불순물을 침적시켜 제염이 효율적이지 않다는 것으로 평가되었다.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2007년도 추계학술대회논문집
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• pp.1386-1390
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• 2007

This paper presents the equipment condition monitoring technology for the process or the equipment using statistical techniques. The equipment condition monitoring system consists of an empirical model to estimate the expected sensor values of process variables and a diagnose model to detect the abnormal condition and to identify the root source of the problem. The empirical model is constructed by the analysis of historic data. The diagnose model uses the sequential probability ratio test (SPRT) technique. The monitoring system was tested with real operating data acquired from the Reactor Coolant Pump Seal in the Nuclear Power Plant. It can detect the system degradation or failure at the early stage since it is able to catch the subtle deviation of process variables from normal condition.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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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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• 제52권10호
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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.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.710-720
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• 1995

원자로에서 펌프에 의해 야기되는 맥동 압력은 원자로 내부 구조물에 진동과 손상을 줄 수 있기 때문에 관심이 증가되고 있다. 본 연구에서는 냉각관과 환형관(원자로 압력 용기와 노심 보호 지지대 사이)으로 구성된 기하 형태에서 펌프에 의해 야기되는 맥동 압력을 해석할 수 있는 수력학적 모델을 개발하였다. 수학적 지배 방정식은 압축성, 비점성 유체에 대해 선형화된 Navier-Stokes 방정식이다. 냉각관과 환형관을 따로 분리하여 해석하고 두영역의 커플링 영향을 고려하였다. 또한 본 기하 형태에서 펌프맥동 압력에 영향을 미치는 주요 기하 인자에 대한 평가를 수행하였다. 본 해석 결과와 실험차를 비교하여 만족할 만한 결과를 얻었다.

• 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.

• 비파괴검사학회지
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• 제9권1호
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• pp.106-110
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• 1989

Bolting degradation problems in primary coolant pressure boundary applications have become a major concern in the nuclear industry. In the bolts concerned, the failure mechanism was either corrosion wastage(loss of bolt diameter) or stress-corrosion cracking.(3) Here the manual ultrasonic testing of RPV(Reactor Pressure Vessel) and RCP(Reactor Coolant Pump) stud has been performed. But it is difficult to detect indications because examiner can not exactly control the rotation angle and can not distinguish the indication from signals of bolt. In many cases, the critical sizes of damage depth are very small(1-2 mm order). At critical size, the crack tends to propagatecompletly through the bolt under stress, Resulting in total fracture.(3) Automatic stud scanner for studs(bolts) was developed because the precise measurement of bolt diameter is required in this circumstance. By use of this scanner, the rotation angle of probe was exactly controlled and the exposure time of radiations was reduced.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 추계학술대회 논문집 학회본부A
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• pp.10-12
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• 1998

Pumping liquid metal in nuclear power plant application by conventional centrifugal pumps pose difficulties such as bearing wear out at high temperatures and leak proof sealing of the liquid metal. MHD machine is obtained by replacing solid conducting secondary of conventional motors with ionized gas or liquid metal. It is used for reactor cooling pump because of construction simplicity, perfect sealing and easy operation/maintenance MHD pump is complicated because it includes electromagnetic and hydrodynamic phenomena. The principle of MHD Pumps is described in this paper. We design small laboratory size Flat Linear Induction Pump(FLIP) for transferring sodium.