• Journal of the Korean institute of surface engineering
• /
• v.51 no.4
• /
• pp.218-223
• /
• 2018

In this study, we proposed a new ultrasonic-chemical decontamination process for decontaminating radioactive corrosion products during the maintenance of reactor coolant pump (RCP). The actual decontamination process was reproduced in the laboratory. And the corrosion characteristics of stainless steel (STS), constituting the RCP interior parts, were examined. The weight-loss measurment and polarization experiment were carried out in order to determine the corrosion characteristics of STS 304 and STS 415 by repeated decontamination processes. The STS 304 presented a little corrosion damage, which was almost indistinguishable from visual observation. The weight-loss rate of STS 304 was also significantly lower. On the other hand, STS 415 showed severe corrosion damage on its surface, greater weight-loss rate and higher corrosion current density than STS 304.

• Journal of Information Technology Services
• /
• v.20 no.3
• /
• pp.1-12
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.17 no.1
• /
• pp.16-24
• /
• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.33 no.6
• /
• pp.418-426
• /
• 2009

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.22 no.8
• /
• pp.1083-1090
• /
• 1998

A kind of three-way check valve, so called hydraulic calve was proposed for the substitute of the density lock of passive reactor such as SPWR (System-Integrated Pressurized Water Reactor). The function of the valve are the separation of the borated water from main coolant loop for normal operation and the insurge of the water into the loop for shutdown and the removal of the decay power when the main coolant flow rate is not enough. To verify the operability and the characteristics of the valve, experimental works were executed with 1/3 scale model calve. Also a diffuser model was proposed for the theoretical analysis of the valve.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2007.11a
• /
• pp.1386-1390
• /
• 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
• /
• 2001.10a
• /
• pp.118-125
• /
• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• The Transactions of the Korean Institute of Electrical Engineers B
• /
• v.49 no.5
• /
• pp.320-326
• /
• 2000

The canned motor of 3-phase induction is used for main coolant pump(MCP). The type of motor is canned-motor that stator and rotor are welded by sealed can. So, cooling water flows in the air gap of the canned motor as an independent cycling cooling system from the air gap to yoke of the motor to prevent high temperature of stator can and to lubricate bearing. Heat exchange is occurred between cooling water in the air gap and cooling water from the exterior pump to prevent rising of temperature in the motor. I has to analyze the characteristics of can exactly because the loss and the heat in the can are very important to design MCP. Therefore, thermal analysis is studied considering the effect of eddy-current los induced in the can.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2014.10a
• /
• pp.110-115
• /
• 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.

• Nuclear industry
• /
• v.27 no.4 s.290
• /
• pp.46-50
• /
• 2007

고리 1호기는 한국에서 최초로 규제 기관에 계속운전을 신청한 원전이다. 2007년 6월에 설계 수명 기간 만료가 되는 고리 1호기는 규제 기관으로부터 계속운전(Continued Operation)에 대한 안전성 심사를 받고 있다. 한수원은 고리 1호기 계속운전 승인을 금년 12월에 받기 위해 최선을 다하고 있으며 지역 주민의 사회적 수용성 확보를 위해 노력중이다. 고리 1호기의 계속운전 기간 동안 안전성을 평가하고 정리한 안전성평가보고서를 한수원은 2006년 6월에 정부에 제출하였다. 고리 1호기는 웨스팅하우스의 2루프 가압경수로이다. 이와 동일한 원전인 일본의 미하마 1,2호기와 겐까이1호기가 계속운전중이며, 미국의 기네이와 포인트 비치 1,2호기가 계속운전 승인을 받았다. 제출한 안전성평가보고서에 대해 한국원자력안전기술원이 심사중이며, 해외 원전과 같이 계속운전을 할 수 있을 것으로 예상하고 있다. 또한 계속운전을 위한 사회적 수용성(Public Acceptance) 확보는 설비의 철저한 안전성 확보 및 지역 주민의 공감대 형성을 통해서 이루어질 것이다. 설계 수명 이후 원자력발전소를 계속 운전하는 것은 이미 선진국에서 시행되고 있다. 2007년 3월 기준으로 미국에서 48기가 운영 허가 갱신 승인을 받았고, 영국은 8기, 일본은 12기가 계속운전중이다. 고리 1호기 성능 지표를 개선시키기 위해서 한수원은 증기발생기, 저압 터빈, 원자로 냉각재 펌프 내장품, 주변압기, 주발전기 등을 교체하였으며, 수명관리 연구, 주기적안전성 평가, 환경 영향 평가를 수행하였다. 2005년 9월에는 미국의 운영 허가 갱신 제도를 참조하여 원자력법이 개정되었다. 이에 한수원은 개정된 원자력법에 맞추어 주기적 안전성평가, 주요 기기에 대한 수명 평가 및 방사능 환경 영향평가를 하였다. 이 세가지 보고서들로 구성된 안전성평가보고서를 2006년 6월에 규제 기관에 제출하였다. 계속운전은 한국을 비롯하여 부존 자원이 부족한 국가들에게는 에너지 자원의 효율적 활용 및 온실 가스 배출을 고려할 때 반드시 필요한 것이다.