• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.1
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• pp.55-62
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• 2016

Nuclear fuels are always exposed to hot temperature and high speed coolant flow during the reactor operation. Thus the fuel rod accompanies small amplitude vibration due to the turbulent flow. The random vibration causes friction between the fuel rod and the grid structure which provides the lateral supports. The friction is critical to the fuel rod fretting wear, and it degrades fuel performance when a severe wear is developed. LuGre friction model is introduced in the paper, and the performance was evaluated comparing to the classical Coulomb model. It is shown that the developed friction force considering the Coulomb friction is not enough to stop or delay the motion while the stick-slip can be simulated using LuGre friction model. Numerical solutions of the two dimensional spacer grid cell model with the modern friction are also reviewed, and it is discussed that the new friction model simulates well the nonlinear mechanism.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4295-4306
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• 2023

The vibration of fuel rods in axial flow is a universally recognized issue within both engineering and academic communities due to its significant importance in ensuring structural safety. This paper aims to thoroughly investigate the stability and nonlinear vibration of a fuel rod subjected to axial flow in a newly designed high temperature gas cooled reactor. Considering the possible presence of thermal expansion and large deformation in practical scenarios, the thermal effect and geometric nonlinearity are modeled using the von Karman equation. By applying Hamilton's principle, we derive the comprehensive governing equation for this fluid-structure interaction system, which incorporates the quadratic nonlinear stiffness. To establish a connection between the fluid and structure aspects, we utilize the Galerkin method to solve the perturbation potential function, while employing mode expansion techniques associated with the structural analysis. Following convergence and validation analyses, we examine the stability of the structure under various conditions in detail, and also investigate the bifurcation behavior concerning the buckling amplitude and flow velocity. The findings from this research enhance the understanding of the underlying physics governing fuel rod behavior in axial flow under severe yet practical conditions, while providing valuable guidance for reactor design.

• Journal of the Korean Society for Nondestructive Testing
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• v.27 no.5
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• pp.432-441
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• 2007

Image processing methods for an extraction of a nuclear fuel rod image and visualization methods of the RGB color data were studied with a sectional image of spent fuel assembly. The fuel rod images could be extracted by using a histogram analysis, an edge detection and RGB rotor data. In these results, a size of the spent fuel assembly could be measured by using a histogram analysis method and a shape of the spent fuel rod could be observed by using an edge detection method. Finally, a various analyses were established for status of the spent fuel assembly by realized various 3D images for the color data in an image of a spent fuel assembly.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.14 no.6
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• pp.512-519
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• 2004

This paper described the free vibration characteristics of Optimized H Type (OHT) spacer grids (SG) supporting the PWR fuel rod. The vibration test and the finite element (FE) analysis are performed under the free boundary condition and the clamped at two points (or three points) in the bottom which is the same one as the experimental condition for the dummy rod continuously supported by spacer grids. A modal test is conducted by the impulse excitation method using an impulse hammer and an accelerometer, and the TDAS module of the I-DEAS software is used to acquire and analyze the sensor signals. The softwares related to the FE analysis are the I-DEAS for the geometrical shape modeling and meshing, and the ABAQUS for solving. The fundamental frequency of the OHT SG by experiment under a clamped condition at two points is 175.18 Hz, and shows a bending mode. We think there is no resonance between the fuel rod and the SG because the SG's frequency is higher than that of the fuel rod existing in the range from 30 to 120 Hz. The fundamental frequency of the SG under the free boundary condition is 349.2 Hz showing a bending mode, and the results between the test and the analysis have a good agreement with maximum 7 ％ in error It is also found that the FE analysis model of the OHT SGs to analyze an impact, a buckling and vibration et al. has been generated with reliability.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.443-449
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• 2014

To check the characteristics of nuclear fuels during an irradiation test, the nuclear fuel rod needs to be disassembled from the test rig located in the pool of the research reactor. Then, the disassembled fuel rod is delivered to the hot cell for intermediate examination. A fuel rod that passes the intermediate examination is delivered to the reactor pool to be reassembled into the test rig. The irradiation test is resumed with the reassembled test rig. Because nuclear fuel rods irradiated by neutrons are highly radioactive, all the disassembly and reassembly processes should be carried out in the pool of the research reactor to prevent operators being exposed to radiation. In particular, because a test rig is 5.4-m long and the reactor pool of HANARO is 6-m deep, special tools need to be developed for performing the disassembly and reassembly processes. In this study, a new assembly design of nuclear fuel rods for intermediate examination is introduced. Furthermore, tools for treating the irradiated fuel rod assembly are introduced, and their performance is verified by an out pile test.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.256-261
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• 2004

To evaluate the effect of spring shape on the fretting wear of nuclear fuel rod, sliding wear tests were performed using three kinds of space grid springs in room temperature air and water. With increasing slip amplitude, wear volume of each spring gradually increased. It is apparently shown that spring with convex shape had a relatively high wear resistance compared with concave shape springs. It is suggested that the ratio of the wear volume to the worn area can be suggested as an efficient and valid parameter to evaluate the wear resistibility of a fuel grid spring.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.257-266
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• 1999

A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.731-736
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• 2000

An optimal design method is adopted for a spacer grid in nuclear power plant. It is made of punched sheet metal process, functioning as springs and dimples supporting fuel rods. For stress analysis of the assembled fuel rod support, a typical cell out of the repeated pattern in the assembly is modeled using 4-node shell elements. A commercial code, ABAQUS, is used for detailed analysis of contacting phenomena with friction. For the optimization, design varibles are taken from geometric parameters representing the shape of the bent leaf spring part and mating contact region with fuel rod. Objective function is considered in relation to mechanical functions and durability. Maximum yon Mises stress is considered in relation to constrained contact stress.

• KSTLE International Journal
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• v.4 no.1
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• pp.31-36
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• 2003

Sliding and impact/sliding wear test in room temperature air and water were performed to evaluate the effect of spring shapes on the wear mechanism of a fuel rod. The main focus was to quantitatively compare the wear behavior of a fuel rod with different support springs (i.e. two concaves, a convex and a flat shape) using a ratio of wear volume to worn area (De)-The results indicated that the wear volumes at each spring condition were varied with the change of test environment and loading type. However, the relationship between the wear volume and worn area was determined by only spring shape even though the wear tests were carried out at different test conditions. From the above results, the optimized spring shape which has more wear-resistant could be determined using the analysis results of the relation between the variation of De and worn surface observations in each test condition.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1819-1824
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• 2004

The common method to improve heat transfer in Nuclear fuel rod bundle is install a mixing vane in space grid. The previous split mixing vane is guides cooling water to swirl flow in sub-channel of fuel assembly. But, this swirl flow decade rapidly after mixing vane and the effect of enhancing the heat transfer vanish behind this short region. The large scale secondary vortex flow was generated by rearranging the inclined angle direction of mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid and the streamwise vorticity in subchannel with LSVF mixing vane sustain two times more than that in subchannel with split mixing vane. The turbulent kinetic energy and the Reynolds stresses generated by the mixing vanes have nearly same scales but maintain twice more than previous type.