• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.591-598
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• 1997

The COBRA/RELAPS code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERA developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• 상하수도학회지
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• 제32권2호
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• pp.107-113
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• 2018

Maintaining adequate residual chlorine concentration is an important criteria to provide secure drinking water. The chlorine decay can be influenced by unstable flow due to the transient event caused by operation of hydraulic devices in the pipeline system. In order to understand the relationship between the transient event and the chlorine decay, the probability density function based on the water demand curve of a hypothetical water distribution system was used. The irregular transient events and the same number of events with regular interval were assumed and the fate of chlorine decay was compared. The chlorine decay was modeled using a generic chlorine decay model with optimized parameters to minimize the root mean square error between the experimental chlorine concentration and the simulated chlorine concentration using genetic algorithm. As a result, the chlorine decay can be determined through the number of transients regardless of the occurrence intervals.

• 한국유체기계학회 논문집
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• 제8권4호
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• pp.20-26
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• 2005

When the pumps are started or stopped for the operation or tripped due to the power failure, the hydraulic transients occur as a result of the sudden change in velocity. The field tests on the waterhammer were carried out for Pangyo booster pumping station in which had six booster pumps and two in-line pumps with the motor of output 1,700 kW, respectively. The booster pumping station was equipped with the pump control valve as the main surge suppression device, and the surge relief valve as auxiliary one. But the pump control valve had not early controlled in the planned closing mode, the slamming occurred to the valve of which abruptly closed during the large reverse flow. Because the positive pressure wave caused by the pump failure was superposed on the slam surge, the upsurge increased so extremely that the pump control valve was damaged. After the air chambers were additionally installed in the booster pumping station, it was preyed that the water supply system acquire the safety and reliability on the pressure surge.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.208-216
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• 2004

The waterhammer occured when the pumps are started or stopped for the operation or tripped due to the power failure, the hydraulic transients occur as a result of the sudden change in velocity. The field tests of the waterhammer were carried out for PanGyo booster pumping station. The PanGyo pumuing station was installed booster pump of 6 sets and in-line pump of 2 sets. The main surge suppression device was equipped with the pump control valve and the surge relief valve as auxiliary. However, the pump control valve had not early controlled in the planned closing mode, and the slamming occurred to the valve of which abruptly closed during the large reverse flow. Because the pressure wave caused by the pump failure was superposed on the slam surge, the upsurge increased so extremely that the shaft of the valve was damaged. After the addition surge suppression device was equipped with air chamber. Further more in-line pump is needed surge suppression device that the pumping station acquired the safety and reliability for the pressure surge.

• 한국유체기계학회 논문집
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• 제7권1호
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• pp.51-57
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• 2004

The waterhammer has recently become more important because the pumping stations were big and the systems conveying the fluid through the large and long transmission pipelines were complex. When the pumps are started or stopped for the operation or tripped due to the power failure, the hydraulic transients occur as a result of the sudden change in velocity As the pressure waves are propagating between the pumping station and the distributing reservoir, the pressure inside the pipe drops to the liquid vapor pressure with the pipeline profile, at which time a vapor cavity forms, and finally the column separation occurs. If the pressure in the pipe is less than the atmospheric pressure, the pipe can be collapsed and destroyed after the water columns separated by the vapor cavity rejoin. During the reverse flow, the pressure is so abnormally increased at the pumping station that the accident of flooding may happen due to the failure of system. In this paper, the field tests on the waterhammer by the startup, stoppage, and power failure of a centrifugal pump were carried out for Yongma transmission pumping station in Seoul. The experimental results were compared with that of the numerical calculations, in which results the procedure of controlled pump normal shut-down and the two-step closing mode of controlling the ball valve for pump emergency stop are proposed to reduce the pressure surge.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.358-369
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• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.702-708
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• 2019

This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5% -100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.

• 대한지하수환경학회지
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• 제7권2호
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• pp.89-96
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• 2000

Slug시험은 수리전도도 예측에 가장 널리 쓰이는 편리한 대수층 실험법이다. Slug시험 중인 관측정 수위 변동은 관측정 반경, 스크린 길이, 다공 매질의 수리전도도와 비저유계수에 영향을 받는다. 이 연구에서는 유한요소기법을 이용한 새 slug시험 모델을 개발하고 그 유용성을 시험하였다. 모델에서 관측정의 수위변동과 다공매질내 지하수유동을 반복기법(iteration technique)을 적용하여, 스크린상의 유동량 측정으로 연결하였다. 이 모델의 수치적 정확도는 Cooper et al. (1967)의 분석해에 대해 검증되었다, 본 방법은 주변 모니터링이 가능한 slug시험의 시뮬레이션, 부분관통과 비저유 계수의 반영 등의 장점이 있다. Slug시험을 통한 다공매질 반영 범위는 비저유계수에 민감하다. 작은 비저유계수일수록 수두압 변화 전파범위가 커져 그 반영률이 증가한다. 관측정의 표준곡선 비교를 통해 비저유계수 예측이 어려우므로, 다공매질 내에서 관측된 수두변화의 표준곡선 활용이 유용할 것이다. 지하수유동의 수직 성분이 커질수록 관측정 수위 회복에 대한 비저유계수의 영향은 더 감소한다. 실제 관측정 주변의 수평-수직한 지하수 유동 해석시, 비저유계수의 무시와 관측데이터의 적용구간 선택에 의한 수리전도도 예측 편차는 거의 무시할 만하며, 수평유동의 경우 분석방법상 편차가 약간 발생한다.