• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1996.04b
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• pp.25-29
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• 1996

The fretting wear characteristics for Zircaloy-4 tube used as fuel rod in the nuclear power plant have been investigated. The fretting wear tester was designed and manufactured for this experiment. This study was focused on main factors of fretting wear, cycle, slip amplitude and normal load. The worn surfaces were observed by SEM.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.1 s.244
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• pp.24-31
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• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In thi present study, the large scale vortex flow(LSVF) is generated by rearranging the 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. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane. Heat transfer in the rod bundle occurs greatly at the same direction to cross flow, and maximum temperature at the surface of bundle drops about 1.5K

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.331-338
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• 1996

In the thermal-hydraulic design, the effect of fuel rod bow is quantified tv the rod bow DNBR penalty which is a key design parameter to assure the coolability of fuel assembly in the pressurized water reactor. In this work, a computer program for the evaluation of the rod bow DNBR penalty based on Westinghouse methodology is developed and its application procedure is proposed. The computer simulation is based on the Monte-Carlo method. The qualification of developed computer program is performed by a comparison of calculational result with that given by Westinghouse's document. A new application procedure is built using batch mean and batch standard deviation. The normality of sample population generated by the batch calculation is confirmed by means of a chi-square test for goodness of fit. On the view point of statistics it is effected that the more reliable design value may be produced by the new application procedure.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.348-348
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• 2001

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3272-3282
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• 2022

To improve the safety and life extension qualities of nuclear fuel rods which is currently made of zirconium (Zr) alloy, research on the application of chromium (Cr) coating was conducted. Cr coating has advantages such as increased corrosion resistance and reduced oxidation rate, but non-destructive thickness evaluation studies are needed to ensure the reliability of the steps taken to provide uniform coating thickness. Eddy current testing (ECT) is a representative non-destructive technique for such as thickness evaluation and surface defect inspection. To inspect changes in thickness at micron scale, the Swept Frequency Eddy Current Testing (SFECT) method was applied to select a frequency range sensitive to changes in thickness. The coating thickness was evaluated using changes in signals, such as that for impedance. In this study, basic research was performed to evaluate the thickness of the Cr coating on a rod using an encircling sensor and the SFECT technique. The sensor design parameters were determined through simulation, after which the new sensor was manufactured. A sensor capable of measuring the thickness of a non-uniformly Cr-coating rod was selected through an experiment evaluating the performance of the manufactured sensor. This was done using the impedance-difference of a Cr-coating rod and a Zr alloy rod. The possibility of evaluation of the Cr coating thickness was confirmed by comparing the experimental results with the selected sensor and the signals of the measured Cr-coating rod. All simulation results were verified experimentally.

• Proceedings of the Korean Operations and Management Science Society Conference
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• 2005.05a
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• pp.106-111
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• 2005

In nuclear power plant, the source of the energy is generated from the nuclear fuel rod. Given a certain level of consumption, the burnt fuel rod should be removed and replaced by a new(fresh) one. The burnt fuel is approximately one third of the whole fuel rods. Currently, this operation is done manually using paper documents and verbal communication and consumes a lot of operation time. In this study, we develop an computerized operation process of nuclear fuel rod replacement procedure based on the ERP(Enterprise Resource planning) methodology.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1377-1380
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• 2008

An irradiation test of a nuclear fuel rod having external and internal tube structure was planned for a performance. To establish fabrication process satisfying the requirements of irradiation test, micro-TIG welding system for fuel rods was developed, and preliminary welding experiments for optimizing process conditions of fuel rod was performed. Fuel rods with 15.9mm diameter and 0.57mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this experiment, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the endcap welding procedure. The soundness of the welds quality of a dual cooled fuel rods has been confirmed by mechanical tests and microstructural examinations.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.41-49
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• 2003

Tube oscillation behaviour is experimentally investigated for the study on the fuel rod fretting that is caused by the flow-induced vibration in nuclear reactor. The experiment was conducted in all at room temperature. The specimen of tube assembly was supported by plate springs which simulated the spacer grids and fuel rods of a fuel assembly. To investigate the influence of contact condition between the grids and rods, normal load of 10 and 5 N, gaps of 0.1 and 0.3 mm were applied. The range of the oscillation at the center of the fuel rod specimen was varied as 0.2, 0.3 and 0.4 mm to simulate the fuel rod vibration due to flow. Displacements near the contact were measured with four displacement sensors during the tube oscillation. As results, the shape of oscillation (phase) varied depending on the contact condition. The oscillation displacement increased considerably from the contact to gap condition. The displacement increased further as the gap size increased. It is regarded that the spring shape influences the tube oscillation behaviour. Simple calculation showed that the slip displacement was very small. Therefore, cumulative damage concept is necessary for the fuel rod wear. The mechanism of plowing is thought required to explain the severe wear in the case of gap existence.