• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.291-296
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• 1998

노내에서 지지격자 스프링의 잔류 탄성변위는 시간(연소도)에 따라 변하게 된다. 이는 격자판의 중성자 조사에 의한 길이방향의 성장으로 지지격자 셀 크기의 증가와 피복관의 크리프에 의한 직경의 감소 및 중성자 조사에 의한 지지격자 스프링력의 이완으로 인한 것이다. 만일 지지격자 스프링의 거동이 변하여 연료봉을 탄성적으로 지지하지 못할 경우 이것은 연료봉의 유체에 의한 진동을 가속시키게 되며, 연료봉과 지지격자 스프링이나 딤플간의 반복적인 고주기의 충격하중은 연료봉의 지지부와 봉간(grid-to-rod)의 프레팅 마모의 원인이 될 수 있다. 따라서 시간에 따라 변하는 변수들의 영향을 고려한 지지격자 스프링의 잔류 탄성변위를 예측할 수 있는 방법론을 정립하여 새로운 지지격자체의 개발시 건전한 연료봉의 지지거동을 평가할 수 있는 도구로 활용하고자 하였다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.4 s.70
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• pp.405-416
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• 2005

This study contains the static buckling tests and static buckling analyses for small size grids and full size grids. The buckling tests and finite element analyses were performed to evaluate the buckling characteristics of the spacer grids in a pressurized water reactor fuel assembly and to evaluate the possibility of the prediction lot the buckling strength of spacer grids. The buckling tests were performed for small size grids and full size grids, and the correlations between buckling strength and the number of straps and the correlations between buckling strength and the number of rows are derived based on the test results. The static buckling analyses were performed to identify the effect of the number of rows and the number of columns on the buckling strength of spacer grid by a finite element method using ANSYS program and the results were compared with the buckling test results.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.7
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• pp.689-695
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• 2012

The friction factor in a rod bundle and the loss coefficient at a spacer grid were examined. As a test section, 25 smooth rods, 9.5 mm in diameter and 2000 mm in length, were prepared and installed in a $5{\times}5$ square array in a square channel. In this case, the P/D (Pitch-to-Diameter ratio) was 1.35. In this work, plain (i.e., no mixing vanes), split-vane, and hybrid-vane spacer grids were tested. In a bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations. Among the spacer grids tested, the hybrid-vane spacer grid presented the largest friction factor in the rod bundle and loss coefficient. This may be because of the flow pattern change induced by large relative plugging of the flow cross section and mixing vane geometry. At Re=$5{\times}10^5$, the predicted loss coefficients of plain, splitvane, and hybrid-vane spacer grids were approximately 0.79, 0.80, and 0.88, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.514-519
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• 1998

범용 전산유체해석(computational fluid dynamics) 코드인 CFX를 이용하여 지지격자 형상에 따른 봉다발 부수로에서의 난류유동 수치해석을 수행하였다 ABB와 SIMENS가 각각 개발한 split vane이 부착된 지지격자와 원자력연구소가 개발중인 회전유동 발생장치가 부착된 지지격자를 포함하는 부수로 난류유동을 분석하였다. 각각의 지지격자 형상에 대해 부수로에서의 축방향 속도, 횡방향 속도, 난류 운동에너지, 와류크기와 압력강하 둥을 비교-분석하였다. 세가지 경우 모두 유사한 경향을 나타냈으나 SIMENS split vane의 유동 전향날개가 크기때문에 와류와 압력강하가 다소 크게 예측되었다. 난류 운동에너지와 와류크기는 지지격자 근처에서 현저히 증가한 후 급격히 감소하는 측정결과를 CFX예측결과에서도 확인할 수 있었다. CFX 예측결과는 지지격자 근처에서 실험 결과와 다소 큰 차이를 보였으나 비교적 부수로 유동특성을 잘 나타낸다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.703-708
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• 1995

지진 및 냉각재상실사고와 같은 외력에 대한 집합체 구조적 건전성 분석을 위한 해석시 필요한 지지격자의 동적 특성치들을 얻기 위해 17$\times$17 JDFA 중간지지격자에 대한 충격시험을 수행하였으며, 지지격자에 대한 허용충격하중값(Crush Strength)을 구하고 Impact Duration Method를 이용하여 집합체 구조해석에 필요한 지지격자의 동적 강성도 (Dynamic Stiffness)를 구하였다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.157-164
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• 2016

The performance of convective heat transfer in rod bundle flow was experimentally evaluated using a twist-vane spacer grid. A $4{\times}4$ square-arrayed rod bundle was prepared as the test section, with a pitch-to-diameter ratio(P/D) of ~1.35. To check the convective heat transfer performance, the circumferential and longitudinal variations in rod-wall temperatures were measured downstream of the twist-vane spacer grid. In the circumferential measurements, the rod-wall temperature toward the twist-vane tip showed the lowest value, which might be due to the deflected water flow caused by the twist-vane. On the other hand, the wall temperature of the longitudinal measurements near the twist-vane spacer grid decreased dramatically, which implies that the convective heat transfer performance was enhanced. A heat transfer enhancement of ~35 % was achieved near downstream of the twist-vane spacer grid, as compared with the upstream value. Based on the present experimental data, a correlation for predicting the heat transfer performance of a twist-vane spacer grid was proposed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.56-71
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• 1996

The local hydrulic characteristics in subchannels of 5$\times$5 nuclear fuel bundles with spacer grids were measured at upstream and downstream of the spacer grid for the investigation of the spacer grid effects on turbulent flow structure by using an LDV(Laser Doppler Velocimeter). The measured parameters are axial velocity and turbulent intensity, skewness factor, and flatness factor. Pressure drops were also measured to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. From these data, it was found that the turbulent mixing and forced mixing occur up to $x/D^h=10$ and 20 from the spacer grid, respectively. The turbulence decay behind spacer grid behaves in the similar decay rate as turbulent flow through mesh grids or screens. Mixing factors useful in subchannel analysis code were correlated from the data and show the highest value near spacer grid and then have a stable values.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.128-128
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.230.1-230
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• 2001