• Nuclear Engineering and Technology
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• 제48권2호
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.1592-1597
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• 2004

Experimental studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel rod bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Experimental studies were carried out at Reynolds Number 60,000 with hydraulic condition. Normal variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. The turbulence generated by split mixing vanes has small length scales so that they maintain only about 10DH after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up $25D_{H}$ after the spacer grid.

• 한국전산유체공학회지
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• 제8권2호
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• pp.8-15
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• 2003

Three nonlinear κ-ε models with the wall function method are applied to the fully developed turbulent flow in a triangular subchannel of a bare rod bundle. Typical predicted quantities such as axial and secondary velocities, turbulent kinetic energy and wall shear stress are compared in details both qualitatively and quantitatively with both each other and experimental data. The nonlinear κ-ε models by Speziale[1] and Myong and Kasagi[2] are found to be capable of predicting accurately noncircular duct flows involving turbulence-driven secondary motion. The nonlinear κ-ε model by Shih et aL.[3] adopted in a commercial code is found to be unable to predict accurately noncircular flows with the prediction level of secondary flows one order less than that of the experiment.

• Journal of Mechanical Science and Technology
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• 제14권8호
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.381-393
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• 1993

이 논문은 하강하는 물의 막과 상승하는 공기유동으로 구성된 수직 다봉채널내서의 역유동 플러딩에 관한 실험결과를 발표한 것이다. 특히 다봉속에 혼합깃이 있는 것과 없는 스페이서 그릿 및 블럭키지가 프러딩의 거동에 미치는 영향이 조사되었다. 시험부에는 5$\times$5 지르칼로이관속이 사용되었다. 플러딩에 관한 전의 해석적 모델과 실험적 상관식들을 본 실험결과와 비교한즉 기존의 모델과 상관식들이 매우 높은 플러딩 곡선을 예측함이 발혀졌다. 스페이서 그릿은 이것이 없는 봉속과 비교하면 낮은 플러딩 공기유량을 나타낸다. 그러나 혼합깃이 있는 스페이서 그릿은 이것이 없는 스페이서 그릿에 비하여 높은 플러딩 공기유량이 요구된다. 블럭키지를 가진 봉속은 더욱더 높은 플러딩 공기유량을 갖는다. 세가지 형식의 시험봉 속에 대한 실험적 플러딩 상관식이 얻어졌다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2008년도 제39회 하계학술대회
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• pp.333-334
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• 2008

Overhead transmission lines are classified by the number of sub conductors per phase. Korean transmission lines use two, four, or six-conductor bundle. Bundle of conductors must have spacers or spacer dampers which keep the proper distance between sub conductors. They can prevent conductors from being vibrated or twisted due to the wind. As for the two-conductor bundle, alternating current flow generates absorption force between sub conductors which may cause a collision of sub conductors. To prevent sub conductors from being vibrated, twisted, and collided, spacer or spacer damper installation method is designed considering vibration characteristics of sub conductors. We have spacer installation method for four or six-conductor bundle lines. However, we don't have it for two-conductor bundle ones. So we have installed spacers at regular intervals in two-conductor bundle lines, and it caused rigid body oscillation of conductors due to the wind which made a flashover between conductors. This paper introduces a vibration characteristic analysis of two-conductor bundle and proposes a spacer installation method for two-conductor bundle lines.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2873-2878
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• 2007

The mixing vanes attached to the spacer grid of rod bundles are used to improve the heat transfer in heat exchanger devices by controlling the characteristics of the flow structures and turbulence. In this study, velocity patterns induced by two types of mixing vane(split and swirl vane) are measured by the PIV technique to better understand how to effect on the cross and secondary vortex flow patterns in $5{\times}$ rod bundle simulating the fuel assembly of the nuclear reactor. A successful measurement of the lateral velocity patterns was conducted using a specially designed beam sheet generator and experimental loop at KAERI. As the result, we found that for the cross flow between subchannels, the split vane is more effective than the swirl vane, while for the secondary vortex flow in each subchannel, the swirl vane's one is larger and longer than split vane's one.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.457-468
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• 2005

An experimental study on post-CHF heat transfer has been performed with a $3\times3$ rod bundle using a vertical steam-water two-phase flow at low flow conditions. The effects of various parameters on the post-CHF heat transfer are investigated and the reasons for the parametric effects are discussed. As the heat transfer regime changes from CHF to post-CHF, the radial wall temperature distribution is changed depending on the pressure and the mass flux conditions. The superheat of the fluid increases considerably with an increase of the wall temperature (or heat flux) and with a decrease of the mass flux. This implies, indirectly, a strong thermal non-equilibrium at high wall temperature and low mass flux conditions. In order to improve the prediction accuracy of the existing post-CHF correlations, it is necessary to perform more experiments, particularly direct measurement of the vapor superheat, and to modify the correlation by considering a strong thermal non-equilibrium at low flow and low pressure conditions.

• 대한기계학회논문집B
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• 제37권10호
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• pp.961-969
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• 2013

본 연구에서는 상용 전산유체역학 소프트웨어인 ANSYS CFX V.13을 사용하여 정상상태, 비압축성, 등온으로 가정된 엇갈림 관 다발 및 일렬 관 다발 주위의 난류유동을 계산하였다. 계산 결과는 전산유체역학 우수사례 지침에 근거하여 격자크기, 대류항 차분법의 정확도 및 난류모델에 대한 민감도 연구에 활용되었고 실험 결과와 정량적으로 비교함으로써 우수사례 지침의 적용성을 평가하였다. 결론적으로 전산유체역학 우수사례 지침이 관 다발 유동 분야에서 상용 전산유체역학 소프트웨어의 예측성능 향상을 반드시 보증하지 않음을 확인하였다.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1905-1915
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• 2019

This paper investigated the influence of simple support girds on flow, irrespective of having mixing vanes, in a 1 × 3 array rod bundle by using CFD methodology and the most accurate turbulence model which could reflect the actual physics of the flow was determined. In this context, a CFD model was created simulating the experimental studies on a single-phase flow [1] and the results were compared with the experimental data. In the first part of the study, influence of mesh was examined. Tetra, hybrid and poly type meshes were analyzed and convergence study was carried out on each in order to determine the most appropriate type and density. k - ε Standard and RSM LPS turbulence models were used in this section. In the second part of the study, the most appropriate turbulence model that could reflect the physics of the actual flow was investigated. RANS based turbulence models were examined using the mesh that was determined in the first part. Velocity and turbulence intensity results obtained on the upstream and downstream of the spacer grid at -3d_h, +3d_h and +40d_h locations were compared with the experimental data. In the last section of the study, the behavior of flow through the spacer grid was examined and its prominent aspects were highlighted on the most appropriate turbulence model determined in the second part. Results of the study revealed the importance of mesh type. Hybrid mesh having the largest number of structured elements performed remarkably better than the other two on results. While comparisons of numerical and experimental results showed an overall agreement within all turbulence models, RSM LPS presented better results than the others. Lastly, physical appearance of the flow through spacer grids revealed that springs has more influence on flow than dimples and induces transient flow behaviors. As a result, flow through a simple support grid was examined and the most appropriate turbulence model reflecting the actual physics of the flow was determined.