• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2002년도 춘계학술대회논문집
• /
• pp.1216-1221
• /
• 2002

The fuel rods in the pressurized water reactor are continuously supported by a spring system called a spacer grid which is one of the main structural components for the fuel rod cluster (fuel assembly). The fuel rods are vibrating within the reactor due to coolant flow. Since the vibration, what is called flow-induced vibration(FIV), can wear away the surface of the fuel rod, it is important to understand the vibration characteristics of it. In this paper, the vibration analyses and the tests for the dummy rods supported by New Doublet(ND) spacer grids are described. A new FE model which reflects the contact area between the rod and ND spacer grid spring is developed to replace the previous one by which a good agreement could not be obtained with the vibration test. The natural frequency and mode shape calculated by both the previous FE model and the new one are compared with those of experiment fur a single-spanned rod supported by two ND spacer grids. The results by the new model show good agreement to experiment as compared with the ones by previous model. In addition, the new FE model is applied to the vibration analysis fur the dummy rod of 2.19 m tall continuously supported by five ND spacer grids. It is also obtained that the analysis results by the new FE model well agree to experiment ones as the single-spanned rod.

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 2006년도 춘계 학술대회 개요집
• /
• pp.39-41
• /
• 2006

The weld quality of spacer grids in Pressurized Water Reactors(PWR) fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and laser beam. The effect of node geometries of spacer grids for GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of GTA and laser beam welded zones have been also compared.

• Nuclear Engineering and Technology
• /
• 제22권4호
• /
• pp.326-336
• /
• 1990

최근 수행된 일련의 실험들을 통하여 핵연료 집합체 지지격자(Spacer Grid)의 존재가 냉각재상실사고시에 핵연료봉으로부터의 열제거에 긍정적인 효과를 미치고 있음이 밟혀졌다. 그 이유는 열원이 없는 지지격자가 연료봉보다 먼저 ？칭이 일어나며 물방울이 지지격자에 부딪쳐서 잘게 부수어져 증발이 쉽게 일어나게 되고 또한 난류효과를 증대시키는 요인이 되기 때문이다. 따라서 냉각재상실사고의 진행 과정에서 첨두피복관온도가 발생하는 재관수 구간의 수면 위쪽에서 유지되는 DFFB에서의 정확한 열전달을 계산하기 위해서는 이들의 고려가 필요하다. 본 논문에서는 DFFB에서 지지격자의 존재로 인해 물방울이 잘게 부수어져 증발이 쉽게 이루어지도록 하는 Droplet Breakup Model을 냉각재상실사고 최적해석 코드인 RELAP5/MOD2에 삽입하였다. 재관수 구간에서 지지격자의 영향을 체계적으로 조사한 FEBA실험에 대해서 검증계산을 수행하여 실험자료와 비교하였다.

• 대한용접접합학회:학술대회논문집
• /
• 대한용접접합학회 2006년 추계학술발표대회 개요집
• /
• pp.125-127
• /
• 2006

A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly for pressurized water reactors(PWRs). The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, weld qualities such as, weld bead size, penetration, spatter, etc. manufactured by various welders were compared and analyzed. Comparison results show that the weld qualities could be improved by selecting the optimal welding condition and also improving the welding technique.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2001년도 추계학술대회논문집A
• /
• pp.355-360
• /
• 2001

Design requirements for the nuclear fuel assembly grid of the pressurized light water reactor(PLWR) are scrutinized from the mechanical/structural point of view. As a result of the scrunity, mechanical/structural test facilities on the spacer grid of the PLWR Fuel are set up in KAERI to find out their mechanical/structural performance.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
• /
• pp.149.2-149.2
• /
• 2005

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
• /
• pp.88-89
• /
• 2019

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2007년도 춘계학술대회B
• /
• pp.2396-2401
• /
• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• 한국전산구조공학회:학술대회논문집
• /
• 한국전산구조공학회 2006년도 정기 학술대회 논문집
• /
• pp.828-835
• /
• 2006

Spacer grid springs support the fuel rods in a nuclear fuel system. The spacer grid is a part of a fuel assembly. Since a spring has repeated contacts with the fuel rod, fretting wear occurs on the surface of the spring. Design is usually performed to reduce the wear. The conceptual design process for the spring is defined by using the Independence of axiomatic design and the design is carried out based on the direction that the design matrix indicates. For detailed design an optimization problem is formulated. In optimization, homologous design is employed to reduce fretting wear. The deformation of a structure is called homologous if a given geometrical relationship holds for a given number of structural points before, during, and after the deformation. In this case, the deformed shape of the spring should be the same as that of the fuel rod. 1bis condition is transformed to a function and considered as a constraint in the optimization process. The objective function is minimizing the maximum stress to allow a local plastic deformation. Optimization results show that the contact occurs in a wide range. Also, the results are verified by nonlinear finite element analysis.

• 한국전산구조공학회논문집
• /
• 제18권4호통권70호
• /
• pp.405-416
• /
• 2005

본 연구에서는 경수로용 핵연료집합체의 전체지지격자(Full Size Grid)와 부분지지격자(Small Size Grid)에 대한 정적 좌굴강도 실험과 전체 지지격자와 부분지지격자를 구성하는 지지격자판(Grid Strap)에 대한 정적 좌굴해석을 수행하여 지지격자의 좌굴특성을 분석하였으며, 분석결과를 이용하여 전체지지격자와 부분지지격자에 대한 좌굴하중값의 예측 가능성을 평가하였다. 좌굴강도 실험은 웨스팅하우스형 연료의 $17{\times}17$셀을 갖는 전체지지격자와 $1{\times}1,\;1{\times}2,\;1{\times}3,\;1{\times}4,\;1{\times}5,\;1{\times}17\;,2{\times}1,\;2{\times}2,\;2{\times}3,\;2{\times}9,\;2{\times}17,\;3{\times}17$ 등의 셀을 갖는 부분지지격자에 대하여 수행하였으며, 실험결과를 이용하여 지지격자의 좌굴강도와 지지격자의 행(rows)과 열(columns) 사이의 관계식을 제시하였다. 좌굴강도 해석은 범용 유한요소해석코드인 ANSYS를 이용하여 수행하였으며, 해석결과를 이용하여 지지격자의 좌굴특성을 분석하고 실험결과와 비교평가 하였다.