• Proceedings of the KSME Conference
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• 2004.04a
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• pp.432-437
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• 2004

KAERI has contrived 16 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 11 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, detailed performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.

• Nuclear Engineering and Technology
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• v.38 no.4
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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.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2396-2401
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• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.7-13
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• 2012

A spacer grid which is one of the most important structural components in a pressurized water reactor fuel is an interconnected array of slotted grid straps, welded at the intersections to form an egg-crate structure. The spacer grid is required to not only protect fuel rods stably but also have sufficient lateral crush strength for the sake of enabling shut-down of the nuclear reactor during abnormal operating environments. Then, the lateral crush strength of the spacer grid is closely related with welding quality of the spacer grid. Previous research on the crush strength analysis of the spacer grid had been performed using only parent material properties. In this study, to investigate the effect on the crush strength of the spacer grid when used mechanical properties in weld zone instead of parent material properties, crush strength analysis considering mechanical properties in weld zone obtained from the instrumented indentation technique was performed and compared the results with the previous research.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.16 no.2 s.107
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• pp.163-168
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• 2006

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle $(5\times5)$ which is called partial fuel assembly is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid to obtain the Flow-Induced Vibration (FIV) characteristics of the scaled fuel assembly over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the assembly prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the scaled test assembly. For the damping factor of the partial fuel assembly and the grid cage at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the partial fuel assembly is about $0.7\%$ with reasonable error of $2\%$ for the previous results. Nonlinear behavior of the partial fuel assembly might be stem mainly from the rod-grid support configuration.

• Proceedings of the KWS Conference
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• 2006.05a
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• pp.39-41
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• 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.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.355-360
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• 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.

• 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.41 no.1
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• pp.53-62
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• 2017

The forced convection heat transfer performance of a twist-vane spacer grid for a dual-cooled annular fuel assembly was examined experimentally. The twist-vane spacer grid was uniquely designed to enhance mixing inside subchannels and mixing between adjacent subchannels. For testing, a $4{\times}4$ square-arrayed rod bundle with narrow gaps between rods was prepared as the dual-cooled annular fuel assembly to be simulated. The pitch-to-rod diameter ratio of simulated dual-cooled annular fuel assembly was 1.08. The experiments were performed under the following conditions: axial bulk velocity, 1.5 m/s and heat flux, $26kW/m^2$. With regard to the circumferential temperature distribution, the lowest rod-wall temperatures upstream and downstream were measured at the subchannel center and the position toward the tip of twist-vane, respectively. With regard to the axial temperature distribution, behind the twist-vane spacer grid, the rod-wall temperature decreased drastically, and the Nusselt number was enhanced by up to 56 %. The present measured data indicate that the twist-vane spacer grid can effectively improve the forced convection heat transfer in the dual-cooled annular fuel assembly with narrow gaps.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.149.2-149.2
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• 2005