Measurements of Void Concentration Parameters in the Drift-Flux Model

윤병조,박군철,정창현,Yun, B.J.,Park, G.C.,Chung, C.H. Korean Nuclear Society 1993 Nuclear Engineering and Technology Vol.25 No.1

가압경수로형 원자로의 정상 비정상 운전시의 열수력학적 거동을 예측하기 위해서는 원자로내기포계수의 분포를 정확히 계산하는 것이 필수적이다. 이러한 기포계수의 정확한 예측을 위하여 많은 모델들이 제시되었다. 이중 drift-flux모델은 그 계산의 정확성과 간결성에 의하여 널리 사용되고 있다. 이러한 drift-flux 모델을 사용하여 보다 더 정확한 기포계수를 예측하기 위해서는 각 상간의 슬립률과 flow regime 에 따른 기포의 운동의 변화가 정확히 고려되어야 한다. Drift-flux 모델에서는 이러한 두 가지 요소가 drift-flux parameter인 $C_{o}$ 와 (equation omitted), 에서 고려된다. 본 연구에서는 이러한 $C_{o}$ 의 실험적 결정을 위하여 원자로 노심을 모사한 4개의 전열봉이 있는 비등이 발생하는 수직사각 유로를 구성하였으며, 완성된 유로내에서 기포계수의 분포 및 기포속도의 분포를 측정하였다. 국부적 기포계수 및 기포속도 분포의 측정에 사용된 방법은 이중탐침법이며 측정이 이루어진 유로내의 유동 상태는 유속이 비교적 느린 low flow rate condition이며 유로내 압력은 3기압 이하이다. 본 실험에서는 액상의 속도는 측정되지 않았으며, 따라서 $C_{o}$ 의 계산을 위하여 (equation omitted)의 실험 상관관계식을 사용하여 유로내 평균 기포계수의 함수로 나타내었다. To predict accurately the thermal hydraulic behavior of light water reactors during normal or abnormal operation, the accurate estimation of the void distribution is required. Up to date, many techniques for predicting void fraction of two-phase flow systems have been suggested. Among these techniques, the drift-flux model is widely used because of its exact calculation ability and simplicity. However, to get more accurate prediction of void fraction using drift-flux model, slip and flow regime effects must be considered more properly In the drift-flux method, these two effects are accounted for by two drift-flux parameters ; $C_{o}$ and (equation omitted). At earlier stage, $C_{o}$ is measured in a circular tube. In this study, $C_{o}$ is experimentally determined by measuring local void fraction and vapor velocity distribution in a rectangular subchannel having 4 heating rods which simulates nuclear subchannels. The measurements are peformed with two-electrical conductivity probes which are known to be adequate for measuring local parameters. The experiments are performed at low flow rate and the system pressure less than 3 atmo spheric pressure. In this experiment, (equation omitted), is not measured, but quoted from well-known empirical correlation to formulate $C_{o}$. Finally, $C_{o}$ is expressed as a function of channel averaged void fraction. fraction.

The Probabilistic Analysis on the Containment Failure by Hydrogen Burning at Severe Accidents in Nuclear Power Plants

박익규,문주현,박군철,Park, I.K.,Moon, J.H.,Park, G.C. Korean Nuclear Society 1994 Nuclear Engineering and Technology Vol.26 No.3

원자력발전소 중대사고시 예상되는 수소생성과 이에 따른 수소연소로 인한 압력증가로 야기되는 격납용기의 파손화률을 몬테카를로 방법을 통하여 계산하였다. 몬테카를로 계산을 수행하기 위해서는 각각의 입력변수들에 대한 적절한 확률분포함수가 요구되는데, 통계적인 처리를 통하여 구하였다. 고리 2호기에 대한 계산을 수행하였으며, 입력변수들에 대한 민감도 분석도 실시하였다. 고리 2호기에서 수소연소로 인한 격납용기의 파손확률은 60% 이하로 계산되었으며, 민감도 분석결과 SFD가 중요한 인자이긴 하지만 다른 인자들도 무시할 수 없는 영향을 미치고 있음이 밝혀졌다. The containment failure probability due to hydrogen burning during severe accidents proceeding in a low pressure sequence is calculated using Monte Carlo method. The probability distribution functions for this Monte Carlo calculation is obtained from the statistical method. The calculations are performed for Kori unit 2, and the sensitivity studies on the input variables-the amount of hydrogen generated at SFD, cerium diameter, cerium length, oxidation rate at FCI, and the amount of hydrogen generated during MCCI-are also performed. It is revealed that SFD is the main factor in hydrogen generation, but the other sources also cannot be neglected. The containment failure probability due to the hydrogen burning lies within 6% in case of Kori unit 2.

단위-셀 실험과 전산유체해석을 통한 블록형 초고온가스로의 노심우회유량 평가

윤수종(S.J. Yoon),진창용(C.Y. Jin),김민환(M.H. Kim),박군철(G.C. Park) 한국전산유체공학회 2009 한국전산유체공학회지 Vol.14 No.2

An accurate prediction of the bypass flow is of great importance in the VHTR core design concerning the fuel thermal margin. Nevertheless, there has not been much effort in evaluating the amount and the distribution of the core bypass flow. ln order to evaluate the behavior and the distribution of the coolant flow, a unit-cell experiment was carried out. Unit-cell is the regular triangular section which is formed by connecting the centers of three hexagonal blocks. Various conditions such as the inlet mass flow rate, block combinations and the size of bypass gap were examined in the experiment. CFD analysis was carried out to analyze detailed characteristics of the flow distribution. Commercial CFD code FLUENT 6.3 was validated by comparing with the experimental results. In addition, SST model and standard k-ε model were validated. The results of CFD simulation show good agreements with the experimental results. SST model shows better agreement than standard k-ε model. Results showed that block combinations and the size of the bypass gap have an influence on the bypass flow ratio but the inlet mass flow rate does not.

미포화 비등 이상유동 내 국소 기포 인자에 대한 실험적 연구

배병언(B.U. Bae),윤병조(B.J. Yun),박원만(W.M. Park),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5

Subcooled boiling test was carried out in the SUBO (Subcooled boiling) test facility to extend a database for a code benchmark. The test section is a vertical annulus with a heater rod at the channel center. For the measurement of local bubble parameters, double optical fiber sensors were applied at the six elevations. Total six test matrixes are chosen for the parametric study of the heat flux, mass flux and inlet subcooling. Void fraction, interfacial area concentration and bubble velocity were measured at 12 radial locations at each elevation. The local bubble parameters show well the characteristics and propagation of a void fraction and an interfacial area concentration along the test section in the subcooled boiling flow. The present data will be suitable for the benchmark, verification and model development for the CFD style codes or existing safety analysis codes.

초고온가스로 원자로 공동냉각계통 상승관의 수직 직사각형 관 내부에서의 난류혼합대류 열전달에 대한 전산유체해석연구

신동호(D.H. Shin),김신엽(S.Y. Kim),김찬수(C.S. Kim),조형규(H.K. Cho),박군철(G.C. Park) 한국전산유체공학회 2017 한국전산유체공학회 학술대회논문집 Vol.2017 No.11

다중블록실험과 전산유체해석을 통한 블록형 초고온가스로의 노심우회유량 평가

윤수종(S.J. Yoon),이정훈(J.H. Lee),김민환(M.H. Kim),박군철(G.C. Park) 한국전산유체공학회 2011 한국전산유체공학회지 Vol.16 No.3

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In ihis regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model for analysis of bypass flow characteristics in detail.

미포화 비등 유동에 대한 Bubble lift-off model 개발 및 1군 계면면적 수송방정식 해석

배병언(B.U. Bae),윤병조(B.J. Yun),윤한영(H.Y. Yoon),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5

The interfacial area transport equation for the subcooled boiling flow was developed with a mechanistic model for the wall boiling source term. It included the bubble lift-off diameter model and lift-off frequency reduction factor model. Those models took into account the bubble's sliding on the heated wall after a departure from the nucleate site and the coalescences of sliding bubbles. To implement the model, the two-phase flow CFD code was developed, which was named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code was validated the experimental data of SUBO (Subcooled Boiling) facility. The computational analysis revealed that the interfacial area transport equation with the bubble lift-off diameter model agreed well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multi-dimensional behavior of void fraction or interfacial area concentration.

원격지 소동력로 (REX10), RTDS 및 SCADA 연계 방안

이상성(S. S. Lee),이송근(S. K. Lee),김광호(K. H. Kim),장길수(G. S. Jang),박군철(G. C. Park),박종근(J. K. Park),문승일(S. I. Moon),윤용태(Y. T. Yoon) 대한전기학회 2010 대한전기학회 학술대회 논문집 Vol.2010 No.11

본 논문에서는 원격지 소동력로 (REX10), RTDS 및 SCADA 연계 방안 및 전력계통 연계를 위한 시뮬레이터를 개발하였다. 또한, REX10을 이용하여 지역에너지시스템을 구현하기 위해서 다양한 부하지역에 대한 시스템 모델 개발, 운전 방식의 모의 및 시뮬레이션을 위하여 하드웨어/소프트웨어 시스템이 개발되었다.

계면면적 수송방정식을 적용한 이상유동 해석코드 개발

배병언(B.U. Bae),윤한영(H.Y. Yoon),어동진(D.J. Euh),송철화(C.H. Song),박군철(G.C. Park) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5

For the analysis of a two-phase flow, the interaction between two phases such as the interfacial momentum or heat transfer is proportional to the interfacial area. So the interfacial area concentration (IAC) is one of the most important parameters governing the behavior of each phase. This study focuses on the development of a computational fluid dynamics (CFD) code for investigating a boiling flow with a one-group IAC transport equation. It was based on the two-fluid model and governing equations were calculated by SMAC algorithm. For checking the robustness of the developed code, the experiment of a subcooled boiling in a vertical annulus channel was analyzed to validate the capability of the IAC transport equation. As the results, the developed code was confirmed to have the capability in predicting multi-dimensional phenomena of vapor generation and propagation in a subcooled boiling.