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소듐고속로 핵연료집합체 측면 오리피스 주입구 난류유동의 전산유체역학 해석
인왕기(W. K. In),정영신(Y. S. Jeong,),이찬(C. Lee),신창환(C. H. Shin),오동석(D. S. Oh),전태현(T. H. Chun),천진식(J. S. Cheon) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11
A liquid sodium is used as coolant in sodium fast reactor(SFR). The reactor coolant pump supplies sodium coolant into a fuel assembly through the inlet chamber with a side orifice. Nine(9) orifices are used in the inlet chamber in three longitudinal and azimuthal directions. This CFD study investigates turbulent flow structure in the inlet chamber and estimate the pressure loss coefficient at the side orifice. The Reynolds number for this CFD simulation is 104, 105 and 2x105 based on the hydraulic diameter and bulk velocity in the orifice. Turbulence models used are the standard k-ε model, SAS-SST model, SSG Reynolds stress model and Large Eddy Simulation(LES). A unsteady flow simulation was performed to more accurately analyze the complex turbulent flow in the inlet chamber with the 9 side orifices. This paper presents the CFD predictions of turbulent flow distribution in the inlet chamber of SFR fuel assembly and the loss coefficient of side orifice.
벽 비등모델을 이용한 과냉비등 유동에 대한 CFD 모의계산에서 벽 인접격자의 영향
인왕기(W.K. In),신창환(C.H. Shin),전태현(T.H. Chun) 한국전산유체공학회 2010 한국전산유체공학회지 Vol.15 No.3
A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and gas(vapour) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 ㎫ and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit for lqiuid phase (y?<SUB>w,l</SUB>) was examined from 101 to 313 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., y?<SUB>w,l</SUB> > 300 at the tube exit.
과냉 비등유동에 대한 CFD 모의계산에서의 벽 인접격자 영향
인왕기(W.K. In),신창환(C.H. Shin),전태현(T.H. Chun) 한국전산유체공학회 2010 한국전산유체공학회 학술대회논문집 Vol.2010 No.5
A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and vapor(steam) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 MPa and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit (y<SUB>w</SUB><SUP>+</SUP>) was examined from 64 to 172 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., y<SUB>w</SUB><SUP>+</SUP> > 100.
봉간격이 좁은 봉다발에서 압력손실의 실험 및 전산해석 평가
신창환(C.H. Shin),이치영(C.Y. Lee),박주용(J.Y. Park),오동석(D.S. Oh),인왕기(W.K. In) 한국전산유체공학회 2011 한국전산유체공학회 학술대회논문집 Vol.2011 No.11
A dual-cooled annular nuclear fuel has been introduced for a significant increase in reactor power. The KAERI has been researching the development of a dual-cooled annular fuel for a power increase in an optimized PWR in Korea, OPR-1000. The pitch-to-diameter ratio of the annular fuel assembly is decreased in order to maintain the fuel amount ratio within the same assembly size as the solid fuel assembly. In the tight lattice rod bundle, a pressure loss for the rod friction may be definitely different from that in the conventional solid fuel assembly. In this study, the friction loss of the 4x4 or 5x5 bare rod bundles without the obstacles such as the spacer grids is measured for the pitch-to-diameter 1.08 and 1.35, respectively. The measured results are compared with the general correlations for the conventional rod bundle. CFD studies are performed to investigate the friction pressure loss for the simulated single rod geometry and the spacer grid effects in the rod bundle is estimated in the view of a friction loss for a rod.
이중냉각핵연료에서 봉단마개의 측면오리피스에 대한 손실계수 평가
곽영균(Y.K Kwack),신창환(C.H. Shin),이치영(C.Y. Lee),박주용(I.Y . Park),전태현(T.R Chun),인왕기(W.K. In) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.11
In the case of complete entrance blockage of the inner channel, the flow rate of the inner channel should be greater than 60% of normal condition in order to avoid a damaged condition of the nuclear fad. For this purpose, the pressure loss coefficient of a side orifice of bottom end plug was evaluated. The pressure loss coefficient were obtained through simulation analysis of single rod for the end plug. According to the blockage rate of the outer channel, flow pattern and pressure loss coefficient of the side hole were investigated
CFD 방법을 이용한 핵연료다발내의 복합 유동혼합 날개의 최적화
인왕기(W. K. In),오동석(D. S. Oh),전태현(T. H. Chun) 대한기계학회 2001 대한기계학회 춘추학술대회 Vol.2001 No.9
The computational fluid dynamics (CFD) method was used to determine an optimum design of hybrid mixing vane in a nuclear fuel bundle. The hybrid mixing vane is a new flow mixing device under development by Korea Atomic Energy Research Institute, which consists of two sets of primary and secondary vanes. Swirling flow and cross flow is primary mechanisms of coolant mixing in the fuel bundle. To maximize the coolant mixing by the hybrid vane, its size and vane angle must be optimized. Assuming an optimum size of the hybrid vane based on engineering judgment, the vane angles, defined as the angle bent from the axial flow direction, were varied from 30˚ to 40˚ and from 25˚ to 45˚ for the primary and secondary vanes, respectively. The swirl mixing increased as both the primary and secondary vane angle increases. The crossflow mixing appeared to increase as the primary vane angle increases and the secondary vane angle decreases. Turbulent mixing showed negligible dependence on the vane angles. Pressure drop due to the hybrid vane continually increased as the vane angle increases. The swirl and cross flow mixing factors were estimated from the predicted velocity distributions in the fuel bundle. The optimal vane angles are judged to be 40˚ and 25˚-35˚ for the primary and secondary hybrid vanes, respectively.
곽영균(Y.K. Kwack),신창환(C.H. Shin),이치영(C.Y. Lee),박주용(J.Y. Park),인왕기(W.K. In) 한국전산유체공학회 2012 한국전산유체공학회 학술대회논문집 Vol.2012 No.5
In nuclear reactors, it is important from the point of view of safety to provide adequate cooling of a fuel. Pressure drop and flow distribution are used as key factors. They were obtained by a numerical study of flow in a plate-type nuclear fuel assembly. In the top end fitting, effects of the sudden transition from the circular pipe to the rectangular duct and the fixing bar presence on the flow distribution in each channel were studied. The pressure drops along the assembly were compared with results of experiments.
박주용(J.Y. Park),신창환(C.H. Shin),이치영(C.Y. Lee),곽영균(Y.K. Kwack),전태현(T.H. Chun),오동석(D.S. Oh),인왕기(W.K. In) 대한기계학회 2012 대한기계학회 춘추학술대회 Vol.2012 No.11
For power uprate of Pressurized Water Reactor(PWR), a dual-cooled annular fuel is being developed in Korea Atomic Energy Research Institute(KAERI). The annular fuel rod is configured to allow the coolant flow through the inner channel as well as outer channel. Since the inner channel is isolated from the neighbor channels unlike the outer channels, an inner channel blockage is one of the principal technical issues of a dual-cooled annular fuel. As a hypothetical event, if the inner channel is completely blocked by the debris, the inner cladding may experience a rapid temperature increase because of no further coolant supply. To complement the entrance blockage of an inner channel a long end plug with side holes is conceptually suggested. The experiment is performed to estimate the flow rate through side holes at the side hole blockage event. In this paper, the inner channel flow rate by the side holes was measured and the loss coefficient of side holes was evaluated.
OPR-1000 원자로의 출력증강을 위한 18×18 원통형 핵연료의 열수력적 타당성 평가
신창환(C. H. Shin),김효일(H. I. Kim),인왕기(W. K. In),전태현(T. H. Chun) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.11
The thermal hydraulic analysis of the 18×18 solid fuel assembly has been carried out for the power uprate of OPR-1000. The suggested 18x18 solid fuel assembly has a structural compatibility for reloading to operating PWR reactors of OPR-1000. In the thermal hydraulic analysis, the mixing effect between the neighboring channels has to be carried out in a subchannel analysis. A subchannel analysis code, MATRA has been developed by KAERI. The main parameters for the thermal hydraulic design, such as a pressure drop and DNBR, and the maximum temperature in a fuel pellet have been estimated. The 18×18 solid fuel in the 120% power uprate showed an increased pressure drop and a similar DNBR behavior. The peak temperature in the fuel centerline, however, was slightly higher than that of the 16×16 solid fuel assembly for the normal operation condition. The peak temperature of a fuel pellet in the solid fuel should be seriously considered for increasing power density.
5X5 핵연료 봉다발에서 지지격자 가공 상태에 따른 압력손실 실험
박주용(J. Y. Park),오동석(D. S. Oh),장석규(S. K. Chang),인왕기(W. K. In) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10
In Korea Atomic Energy Research Institute(KAERI), it holds a patent right in the U.S.A, Korea, and Japan about development of fuel rod spacer grid for high performance of Pressurized Water Reactor(PWR). We performed experiments about hydraulic characteristics of the developed spacer grid. There were 4 types of spacer grid such as Plain Spacer Grid, Chamfering/Coining, spacer grid with mixing vane without Chamfering/Coining treatment(Mixing Vane), spacer grid with spot welding and chamfering/coining treatment(Mixing Vane_SC), and spacer grid with line welding in Mixing Vane_SC(Mixing Vane_LC). We compared pressure drop coefficient and found the pressure drop coefficient is greatly affected by chamfering/coining treatment and wasnt affected by welding procedure.