• Corrosion Science and Technology
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• v.18 no.5
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• pp.175-181
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• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2006.05a
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• pp.141-144
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• 2006

In this study, we developed a decladding device which separates 250 mm length of simulated nuclear spent fuel rod into the pallets and the pieces of the hulls after inserting the rod cut into the module with several pairs of blades. To improve the performance of the equipment, we considered some mechanisms to prevent the rod cut from being exposed or bounced into the hot-cell, to reduce the operation time, and to insert the rods automatically. It is expected that the newly developed system will contribute to prevent radioactive pollution in the hot-cell, reduce the operation time, and to increase the safety of the operators. As a result of the performance test for some mockup fuel rod cuts in the ACP(Advanced Spent Fuel Control Process) facility, it was verified that the decladding device could be applied to the actual fuel rod cut. And it will be able to use for a scale-up facility in the future.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.17 no.9
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• pp.819-825
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• 2007

Vibration characteristics of an duel-cooling cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1997.10a
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• pp.177-183
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• 1997

The fuel rod for PWR is supported by the spring of the sapcer grid to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in the reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. Here considering the spring behaviour as a function of burnup the reaction forces of the springs are calculated by the finite element program developed herein to evaluate the integrity of the fuel rod from fretting. The results are compared with previous data and ANSYS for the validation of the program and procedures.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.13 no.5
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• pp.384-392
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• 2003

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 vibrate within the reactor due to coolant flow. Since the vibration, which is called flow-induced vibration(FIV) can wear away the surface of the fuel rod, it is important to understand it's vibration characteristics. 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 for a single-spanned rod supported by two ND spacer grids. The results of the new model showed good agreement with the experiment compared with those of previous model. In addition. the new FE model is applied to the vibration analysis for the dummy rod of 2.189 mm tall continuously supported by five ND spacer grids. It is also obtained that the analysis results of the new FE model well agreed to experiment ones as the single-spanned rod.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.05a
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• pp.1216-1221
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• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.2
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• pp.201-206
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• 2011

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a pressurized water reactor (PWR), and it affects the performance of the fuel assembly. The primary design requirement is that the mechanical integrity of the fuel rod should be maintained by the spacer grid assembly during the operation of the reactor. It was known that fretting damage to the fuel rod can be reduced by adjusting the relative moving displacement between the fuel rod and its support. In this study, we used the finite element method to evaluate the characteristics of a sliding support designed to reduce fretting damage of fuel rods.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1859-1863
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• 2007

A new type of a fretting wear tester has been designed and developed in order to simulate the actual vibration behavior of a nuclear fuel rod for springs/dimples in room temperature. When considering the actual contact condition between fuel rod and spring/dimple, if fretting wear progress due to the flow-induced vibration (FIV) under a specific normal load exerted on the fuel rod by the elastic deformation of the spring, the contacting force between the fuel rod and dimple that were located in the opposite side should be decreased. Consequently, the evaluation of developed spacer grids against fretting wear damage should be performed with the results of a cell unit experiments because the contacting force is one of the most important variables that influence to the fretting wear mechanism. Therefore, it is necessary to develop a new type of fretting test rig in order to simulate the actual contact condition. In this paper, the development procedure of a new fretting wear tester and its performance were discussed in detail.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.175-180
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• 1997

It has been known that eater-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel dadding tube. The fuel dad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide later thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-destructive Test) hot-cell or PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness ore predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test us peformed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic result obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1997.10a
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• pp.630-633
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• 1997

Spent fuel decladding device and dry voloxidizer is to separate the spent pellet from spent fuel rod cut by 250mm and to convert the spent pellet into powder form for reuse and/or disposal of the spent fuel. There are two methods in decladding and voloxidation of spent fuel, that is, wet method with chemical material and dry method with mechanical device. In this study, to examine the fuel rod decladding process and the pellet voloxidation process, the devices for the spent fuel decladding and the pellet voloxidation with dry method are developed. The decladding machine is designed to separate pellets from fuel rod by slitting device. And, the voloxidizer is designed to convert the spent pellet which is ceramic form into powder form by oxidation using the multi step mesh, vibrator, and air in the high temperature environment. The result of this study, such as operation condition et., will be utilized in the design of the machine for demonstration.