• Nuclear Engineering and Technology
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• 제42권3호
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• pp.279-296
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• 2010

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.501-506
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• 1997

This paper presents the hydraulic characteristics measured by using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops fer each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regard ins the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented.

• Nuclear Engineering and Technology
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• 제24권3호
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• pp.263-273
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• 1992

핵연료 집합체의 속도분포, 압력강하는 열수력 설계와 안전해석에 중요하다. 본 실험적 연구의 목적은 봉다발 지지 격자 하류에서의 수력학적 혼합을 고찰하는데 있다. 이 연구에서 가압경수로형 5X5 봉다발 부수로의 상세한 수력학적 특성들을 1차익 He-Ne LDV를 이용하여 측정하였다. 축방향 유속, 난류강도와 압력강하를 주로 측정하였고 LDV의 정렬을 조정하여 측방향의 유속, 난류강도, Reynolds 전단응력 등도 역시 측정하였다. 봉다발의 마찰계수와 지지 격자의 손실계수는 측정된 압력 강하로부터 평가하였다. 서로 다른 종류의 지지 격자의 수력학적 혼합성능을 이웃하는 부수로 간에서의 난류 횡류 혼합률을 예측함으로써 고찰할 수 있었다.

• Nuclear Engineering and Technology
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• 제28권2호
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• pp.162-174
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• 1996

서로 다른 지지격자들이 인접한 6$\times$6 핵연료 봉다발부수로내에서 국부 수력특성인자들을 레이저 유속 측정 장치인 LDV(Laser Doppler Velocimeter)를 이용하여 측정하였다. 6$\times$6 봉다발은 서로 다른 지지격자를 가진 3$\times$6 봉다발이 서로 인접하여 이룬 형상이다. 본 연구에서는 다른 형상과 다른 수력저항을 갖는 지지격자간들의 열수력적 상호작용을 규명하는데 그 목적이 있다. LDV를 이용하여 축방향 및 횡방향 속도, 난류강도 등의 측정 인자들을 측정하였다. 또한 압력강하를 측정하여 지지격자의 손실계수와 봉다발의 마찰계수를 구하였다. 수력실험결과에 근거하여 지지격자에 기인된 열혼합현상에 관한 것을 연구하였다. DNB의 정성적인 기준이라고 할 수 있는 swirl인자를 정의하고 횡방향속도 실험인자로부터 구하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.95-101
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• 1996

In this paper, the steady 2-dimensional model for a long horizontal line with different end temperatures undergoing natural convection at very high Rayleigh number is proposed to numerically investigate the heat transfer and flow characteristics. The dimensionless governing equations are solved by using SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm which is developed using control volumes and staggered grids. The numerical results are verified by comparison with the operating PWR test data. The analysis focuses on the effects of variation of the heat transfer rates at the pipe surface, the thermal conductivities of the pipe material and the thickness of the pipe wall on the thermal stratification. The results show that the heat transfer rate at the pipe surface is the controlling parameter. A significant reduction and disappearance of thermal stratification phenomenon is observed at the Biot number of 5.0$\times$10$^{-2}$. The results also show that the increment of the thermal conductivity and thickness of the wall weakens the thermal stratification and somewhat reduces azimuthal temperature gradient in the pipe wall. Those effects are however minor, when compared with those due to the variation of the heat transfer rates at the surface of the pipe wall.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.377-382
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• 1995

An evaluation method of spacer grid thermal mixing performance in rod bundles is suggested based on hydraulic tests in a single phase flow. Heat transfer correlation was derived by the analogy between momentum and heat transfer. Three of major factors, such as blockage ratio of spacer grid, convective flow swirling, and turbulent intensity, were found to be significantly influential to the spacer grid thermal mixing performance. Local heat transfer near spacer grid was predicted for the hydraulic test of 6 ${\times}$ 6 rod bundles with neighboring different spacer grids.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1386-1395
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• 2020

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.