• Korean Journal of Materials Research
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• v.10 no.6
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• pp.422-429
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• 2000

To investigate the effect of annealing temperature and time on the recrystallization behavior and microstructure of Zr-based alloys, the specimens of Zr-0.8Sn-0.4Nb-0.4Fe-0.2Cu, Zr-1Nb, Zircaloy-4, and unalloyed Zr were cold-worked and annealed at 400, 500, 600, 700, 800, $900^{\circ}C$ for 30 to 5000 minutes. The hardness, microstructure and precipitate of the specimens were investigated by using micro-hardness tester, optical microscope and transmission electron microscope, respectively. The recrystallization of Zr-based alloys occurred between $400^{\circ}C$ and $600^{\circ}C$. As the content of alloying elements increased, the hardness and recrystallization temperature of the alloys increased though the grain sizes after recrystallization decreased. It was supposed that the hardness of Zr-based alloy with Fe or Cu increased during recovery by the formation of Fe or Cu precipitates.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.1
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• pp.67-75
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• 2013

The flow stress of zircaloy-4 used in the spacer grid supporting a nuclear fuel rod was determined by the Johnson-Cook model, and model parameters were determined using reverse engineering. Parameters such as A, B, n and $\dot{\varepsilon}_0$ were determined by the tensile test result. To obtain the parameters C and m, a slot milling test and numerical simulation were performed. The objective functions were defined as the difference between the experimental and the simulation results, and then, the parameters were determined by minimizing the objective function. To verify the validity of the determined parameters, cross-verification for each case was conducted through a shearing test and simulation. The results tend to show agreement with the experimental results, such as the features of sheared edges and maximum punch force, with the correlation coefficients exceeding at least 0.97.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2041-2046
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• 2020

The present article investigates the influence of hydrogen concentration on the creep performance of cold-worked stress-relieved unirradiated Zircaloy-4 cladding tube using nanoindentation technique. The as-received Zircaloy-4 tube is hydrided to the concentrations of 600 ppm and 900 ppm using gaseous hydrogen charging method. Constant load indentation creep tests are performed for a dwell period of 600 s in the temperature range of 300℃-500 ℃ at 1000 μN, 2000 μN, and 3000 μN. The impact of hydrogen is evaluated in terms of steady state power law creep exponent and activation energy. The power law creep exponent decreases with increase in hydrogen concentration, however, it remains fairly constant with increase in temperature up to 500 ℃. Moreover, activation energy too decreases significantly with increase in hydrogen concentration. The mean stress exponent and activation energy are found to be 3.58 and 28.67 kJ/mol, respectively, for as-received sample.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.88-95
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• 2002

Fretting of fuel rod cladding material, Zircaloy-4 Tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to bean ideal solution to fretting damage since fretting is closely related to wear, corrosion and fatigue. Therefore, in this study the fretting wear experiment was peformed using TiAIN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaioy-4 tube as one of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAIN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and the fretting wear mechanisms were delamination and plastic flow following by brittle fracture at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher slip amplitude.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.739-744
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• 1996

The U solubility in the Zircaloy melt including the other investigators' result was investigated in a range of reaction temperatures from 2223k to 2473k and for the mole ratios of UO$_2$ crucible/Zircaloy-4 melt(subsequently abbreviated as UO$_2$/Zry) from 2.4 to 18.2, The U solubility in the melt increased with increasing reaction temperature and with decreasing the mole ratio of UO$_2$/Zry. An empirical correlation was obtained as functions of UO$_2$/Zry mole ratio and reaction temperature including other investigators' results. The experimental results with use of internally heated fuel element simulators were analyzed by the empirical correlation from UO$_2$ crucible experiments.

• Tribology and Lubricants
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• v.25 no.4
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• pp.250-255
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• 2009

Oxidation characteristics of Zirlo and Zircaloy-4 tubes, which are widely used as nuclear power fuel cladding, are studied in steam environment up to $1200^{\circ}C$. Oxidation resistances are compared in terms of the mass increase due to the absorption of oxygen. The evolution of microscopic structure accompanied with the oxidation process is investigated. Also, the influence of oxidation on the fretting wear characteristics of the tubes is studied. Piezo-electrically actuated rig is employed to fret the tubes with cross-contacting arrangement. Wear scar is observed and measured, by using microscopes and a 3D-profiler. The results of fretting wear are quantified in terms of scar size, wear volume and wear coefficient, and compared for the three different tube materials of oxidated Zirlo, virgin Zirlo and Zircaloy-4.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.2
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• pp.231-238
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• 2003

The fuel of light water reactor is used for several years under high temperature and pressure, so it needs to be clad with high corrosion resistance material. The cladding materials must have the characteristics of low absorption of a neutron and high corrosion resistance. Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor have been used as cladding materials and Zirlo has been developed as the material for preventing the corrosion. If the fracture of the cladding tube occurs during operation, it will cause the economic loss to shut down and replace the system. So it is needed to evaluate the integrity of the cladding materials. In this paper, the tensile characteristics of the cladding materials were investigated for the basic research of fracture characteristics. Also the residual stress was analyzed to compare the tube type(original type) specimen and the flattened type specimen.

• Proceedings of the KWS Conference
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• 2009.11a
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• pp.33-33
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• 2009

This study is to develop the resistance welding apparatus and investigate the welding characteristics of Zircaloy-4 end-plate of fuel bundle in the cases of the resistance welding and the laser beam welding. The welding parameters which affect the weld nugget and the torque values have been also compared. The effect of the torque strength of end-plate welding using by the resistance welding and the laser beam welding has been studied and optimum conditions of Zircaloy-4 end-plate welding have been found. Futhermore, micro-structures and micro-hardness of the resistance welded specimens have been also compared.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.489-494
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• 2001

The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

• Journal of Welding and Joining
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• v.20 no.6
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• pp.72-72
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• 2002

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and find the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test(400℃), the fracture is not happened in the welding part but base metal and the result of corrosion test(400℃ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.