• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1119-1126
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• 2021

Spacer grid with mixing vane had been widely used in nuclear reactor core. One of the main feather of spacer grid with mixing vane was that strong swirl flow was formed after the spacer grid. The swirl flow not only changed the bubble generation in the near wall field, but also affected the bubble behaviors in the center region of the subchannel. The interaction between bubble and the swirl flow was one of the basic phenomena for the two phase flow modeling in fuel assembly. To obatin better understanding on the bubble behaviors in swirl flow, full three dimension numerical simulations were conducted in the present paper. The swirl flow was assumed in the cylindral calculation domain. The bubble interface was captured by Volume Of Fluid (VOF) method. The properties of saturated water and steam at different pressure were applied in the simulation. The bubble trajectory, motion, shape and force were obtained based on the bubble parameters captured by VOF. The simulation cases in the present study included single bubble with different size, at different angular velocity conditions and at different pressure conditions. The results indicated that bubble migrated to the center in swirl flow with spiral motion type. The lateral migration was mainly related to shear stress magnitude and bubble size. The bubble moved toward the center with high velocity when the swirl magnitude was high. The largest bubble had the highest lateral migration velocity in the present study range. The effect of pressure was small when bubble size was the same. The prelimenery simulation result would be beneficial for better understanding complex two phase flow phenomena in fuel assembly with spacer grid.

• 대한기계학회논문집B
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• 제36권7호
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• pp.689-695
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• 2012

지지격자가 있는 봉다발과 축방향으로 평행한 유동에서, 봉다발 마찰계수와 지지격자 손실계수를 평가하였다. 시험부는 외경 9.5 mm, 길이 2000 mm 인 봉 25 개를 $5{\times}5$ 정사각 구조로 배열하여 제작하였으며, 봉 중심간 거리와 봉 외경의 비는 1.35 였다. 지지격자로는 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자를 이용하였다. 지지격자가 없는 봉다발의 마찰계수는 기존 상관식과 비교적 잘 일치하였다. 지지격자가 있는 봉다발 실험의 경우, hybrid-vane 지지격자에서 봉다발 마찰계수 및 지지격자 손실계수가 가장 크게 측정되었으며, 이는 지지격자의 유동단면 막음비 증가와 혼합날개 형상에 의한 유동 교란이 증가되기 때문인 것으로 판단된다. Re=$5{\times}10^5$ 조건에서 plain 지지격자, split-vane 지지격자, hybrid-vane 지지격자의 손실계수는 약 0.79, 0.80, 0.88 로 예측되었다.

• Nuclear Engineering and Technology
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• 제33권5호
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• pp.547-555
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• 2001

In pressurized light water reactor fuel assembly, spacer grids support nuclear fuel rods both laterally and vertically. The fuel rods are supported by spacer grid springs and grid dimples that are located in the grid cell. The support system allows for some thermal expansion and imbalance of the fuel rods. The imbalance is absorbed by elastic energy to prevent coolant flow- induced vibration damage. Design requirements are defined and a design process is established. The design process includes mathematical optimization as well as practical design method. The shape of the grid spring is designed to maintain its function during the lifetime of the fuel assembly. A structural optimization method is employed for the shape design. Since the optimization is carried out in the linear range of finite element analysis, the optimum solution is verified by nonlinear analysis. A good design is found and the final design is compared with the initial conceptual design. Commercial codes are utilized for structural analysis and optimization.

• 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.

• International Journal of Air-Conditioning and Refrigeration
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• 제9권2호
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• pp.77-84
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• 2001

Experimental and computational studies were carried out to investigate the turbulent heat transfer enhancement of the cooling system in nuclear reactor by large scale vortex generation. The large scale vortex motion was generated by rearranging the inclination angels of mixing vanes to the coordinate direction. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity concept based on $\kappa{-}\varepsilon$ model was employed to calculate the turbulent heat and momentum transfers in the subchannel. The turbulences generated by split mixing vanes has small length scales so that they maintain only about $10D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex motions continue longer and remain up to $25D_H$ after the spacer grid.

• 대한기계학회논문집B
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• 제41권1호
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• pp.53-62
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• 2017

이중냉각 환형핵연료 집합체를 위한 비틀림 혼합날개 지지격자의 강제대류열전달 성능을 실험적으로 평가하였다. 비틀림 혼합날개 지지격자는 부수로 간 혼합뿐 아니라 부수로 내 혼합을 동시에 증대시킬 수 있도록 설계되었다. 실험을 위한 이중냉각 환형핵연료 모의 집합체로, 봉 중심 간 거리와 봉 외경의 비가 1.08인 봉 간격이 좁은 $4{\times}4$ 정사각 배열의 봉다발을 준비하였다. 실험은 봉다발 유동의 축방향 평균속도가 1.5 m/s, 열유속은 $26kW/m^2$인 조건에서 수행하였다. 원주방향 온도 분포의 경우, 지지격자 상류에서는 부수로 중심 벽면에서, 하류에서는 비틀림 혼합날개 끝이 향하는 벽면에서 온도가 가장 낮게 나타났다. 축방향 온도 분포의 경우, 지지격자 하류 근처에서 온도가 급격하게 감소하는 것으로 측정되었고, 비틀림 혼합날개에 의해 누셀트 수는 최대 56 % 증대되는 것으로 나타났다. 본 실험결과를 토대로 봉 간격이 좁은 이중냉각 환형핵연료 집합체에서 비틀림 혼합날개 지지격자에 의해 강제대류열 전달 성능이 효과적으로 증대될 수 있음을 확인하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.1373-1376
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• 2008

A supporting structure for a long tube bundle of a large diameter is considered in this paper. The primary purpose of the present study is to develop a spacer grid structure for a so-called "dual cooled nuclear fuel", which has been being studied for a nuclear power uprate. The outer diameter of the fuel rod increases considerably from the conventional one. So a completely new shape of the supporting structure (spacer grid) needs to be developed. One of the challenges is to insert a supporting tube into the cross points of the grid straps. To meet a supporting performance, the load vs. displacement characteristics should be obtained. So the present study focuses on the finite element analysis technology to evaluate the characteristics through a parametric study. As a result, major influencing parameters are investigated for an optimized spacer grid design.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.173.1-173.1
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• 2005

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.375-382
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• 2006

Characterization tests (load vs. displacement curve) are conducted for the springs of Zirconium alloy spacer grids for an advanced LWR fuel assembly. Twofold testing is employed: strap-based and assembly-based tests. The assembly-based test satisfies the in situ boundary conditions of the spring within the grid assembly. The aim of the characterization test via the aforementioned two methods is to establish an appropriate assembly-based test method that fulfills the actual boundary conditions. A characterization test under the spacer grid assembly boundary condition is also conducted to investigate the actual behavior of the spring in the core. The stiffness of the characteristic curve is smaller than that of the strap-wised boundary condition. This phenomenon may cause the strap slit condition. A spacer grid consists of horizontal and vertical straps. The strap slit positions are differentiated from each other. They affords examination of the variation of the external load distribution in the grid spring. Localized legions of high stress and their values are analyzed, as they may be affected by the spring shape. Through a comparison of the results of the test and FE analysis, it is concluded that the present assembly-based analysis model and procedure are reasonably well conducted and can be used for spring characterization in the core. Guidelines for improving the mechanical integrity of the spring are also discussed.

• 설비공학논문집
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• 제12권11호
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• pp.1004-1011
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• 2000

Experimental and computational studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity heat flux model and $k-varepsilon$ model were employed to analyze the turbulent heat and fluid flows in the subchannel. The turbulence generated by split mixing vanes has small length scales so that they maintain only about $10 D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up to $25 D_H$after the spacer gird.