• 대한기계학회논문집B
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• 제40권3호
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• pp.157-164
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• 2016

봉다발 유동에서 비틀림 혼합날개 지지격자의 대류열전달 성능을 실험적으로 평가하였다. 시험부는 $4{\times}4$ 정사각 배열의 봉다발로 제작하였고, 모의 봉다발에서 봉 중심 간 거리와 봉 외경의 비는 ~1.35이다. 대류열전달 성능 평가를 위해 혼합날개 지지격자 하류에서 봉 벽면 온도의 원주방향 및 축방향 분포를 측정하였다. 원주방향의 경우, 지지격자 하류에서 비틀림 혼합날개 끝이 향하는 벽면의 온도가 가장 낮게 나타났는데 이는 비틀림 혼합날개에 의해 왜곡된 유동 때문으로 판단된다. 반면, 축방향의 경우, 혼합날개 지지격자 근처에서 벽면의 온도가 크게 낮아졌는데, 이는 비틀림 혼합날개에 의해 대류열전달이 향상됨을 의미한다. 비틀림 혼합날개 지지격자에 의해 대류열전달 성능은 지지격자 상류에 비해 지지격자 근처 하류에서 ~35 % 향상되었고, 실험데이터를 기반으로 비틀림 혼합날개 지지격자에 대한 열전달 성능 예측 상관식을 제안하였다.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.598-603
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• 2003

A CFD analysis has been performed to investigate turbulent heat transfer in a triangular rod bundle with a pitch-to-diameter ratio(P/D) of 1.06. Anisotropic turbulence models predicted the turbulence-driven secondary flow in the triangular subchannel and the distributions of time mean velocity and temperature showing significantly improved agreement with the measurements over the linear standard ${\kappa}-{\varepsilon}$. The anisotropic turbulence models predicted turbulence structure in large flow region fairly well but could not predict the very high turbulent intensity of azimuthal velocity observed in narrow flow region(gap).

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.1819-1824
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• 2004

The common method to improve heat transfer in Nuclear fuel rod bundle is install a mixing vane in space grid. The previous split mixing vane is guides cooling water to swirl flow in sub-channel of fuel assembly. But, this swirl flow decade rapidly after mixing vane and the effect of enhancing the heat transfer vanish behind this short region. The large scale secondary vortex flow was generated by rearranging the inclined angle direction of mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid and the streamwise vorticity in subchannel with LSVF mixing vane sustain two times more than that in subchannel with split mixing vane. The turbulent kinetic energy and the Reynolds stresses generated by the mixing vanes have nearly same scales but maintain twice more than previous type.

• 설비공학논문집
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• 제18권10호
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• pp.811-818
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• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In the present study, the large scale vortex flow (LSVF) is generated by rearranging the mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about $35D_h$ after the spacer grid. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane.

• 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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• 제10권2호
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• pp.79-86
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• 1978

본 연구의 목적은 수직 연료봉 집합체에서 물-공기 2상 유동일 경우 공기분포현상에 관한 실험적 데이타를 얻는데 있다. test-section은 6각형, 61개의 연료봉 집합체로 구성되며, 각 연료봉은 helical spacers로 감겨져 있고, 사용되는 유체는 공기와 물이다. 실험은 크게 2부분으로 나누어서 물의 유량을 일정하게 하고 공기의 유량을 증가시킬 경우와 물자 공기의 유량을 동시에 증가시킬 경우의 공기분포현상에 관해 실시하였다. 공기는 4구멍을 통해 각각 주입시켰다. 보이드율의 측정은 전기적 Void-needle 방법을 적용하였으며 그 결과는 도표를 통해 보여주고 있다. 이 실험의 결과로써 물의 유랑을 증가시킬 수록 공기분포는 균일하게 되며, 공기 공급 위치는 공기분포에 큰 영향을 미치고 있음이 입증되었다.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4357-4366
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• 2023

MARS-KS, a domestic regulatory confirmatory code of Republic of Korea, had been developed by integrating RELAP5/MOD2 and COBRA-TF. The integration of COBRA-TF allowed to extend the capability of MARS-KS, limited to one-dimensional analysis, to multi-dimensional analysis. The use of COBRA-TF was mainly focused on subchannel analyses for simulating multi-dimensional behavior within the reactor core. However, this feature has been remained as a legacy without ongoing maintenance. Meanwhile, MARS-KS also includes its own multidimensional component, namely MULTID, which is also feasible to simulate three-dimensional convection and diffusion. The MULTID is capable of modeling the turbulent diffusion using simple mixing length model. The implementation of the turbulent mixing is of importance for analyzing the reactor core where a disturbing cross-sectional structure of rod bundle makes the flow perturbation and corresponding mixing stronger. In addition, the presence of this turbulent behavior allows the secondary transports with net mass exchange between subchannels. However, a series of assessments performed in previous studies revealed that the turbulence model of the MULTID could not simulate the aforementioned effective mixing occurred in the subchannel-scale problems. This is obvious consequence since the physical models of the MULTID neglect the effect of mass transport and thereby, it cannot model the void drift effect and resulting phasic distribution within a bundle. Thus, in this study, the turbulence mixing model of the MULTID has been improved by means of the inter-channel mixing model, widely utilized in subchannel analysis, in order to extend the application of the MULTID to small-scale problems. A series of assessments has been performed against rod bundle experiments, namely GE 3X3 and PSBT, to evaluate the performance of the introduced mixing model. The assessment results revealed that the application of the inter-channel mixing model allowed to enhance the prediction of the MULTID in subchannel scale problems. In addition, it was indicated that the code could not predict appropriate phasic distribution in the rod bundle without the model. Considering that the proper prediction of the phasic distribution is important when considering pin-based and/or assembly-based expressions of the reactor core, the results of this study clearly indicate that the inter-channel mixing model is required for analyzing the rod bundle, appropriately.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Journal of Mechanical Science and Technology
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• 제20권9호
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• pp.1514-1524
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• 2006

This work presents the three-dimensional analysis of flow and heat transfer performed for a wire-wrapped fuel assembly of liquid metal reactor using Reynolds-averaged Wavier-Stokes analysis in conjunction with 557 model as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary conditions at inlet and outlet of the calculation domain. Three different assemblies, two 7-pin wire-spacer fuel assemblies and one bare rod bundle, apart from the pressure drop calculations for a 19-pin case, have been analyzed. Individual as well as a comparative analysis of the flow field and heat transfer have been discussed. Also, discussed is the position of hot spots observed in the wire-spacer fuel assembly. The flow field in the subchannels of a bare rod bundle and a wire-spacer fuel assembly is found to be different. A directional temperature gradient is found to exist in the subchannels of a wire-spacer fuel assembly Local Nusselt number in the subchannels of wire-spacer fuel assemblies is found to vary according to the wire-wrap position while in case of bare rod bundle, it's found to be constant.