• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1250-1258
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• 2021

In this paper, the mechanical performance of cracked surface-coated Zircaloy cladding, which has different coating materials, coating thicknesses and initial crack lengths, has been investigated. By analyzing the stress field near the crack tip, the safety zone range of initial crack length has been decided. In order to determine whether the crack can propagate along the radial (r) or axial (z) directions, the energy release rate has been calculated. By comparing the energy release rate with fracture toughness of materials, we can divide the initial crack lengths into three zones: safety zone, discussion zone and danger zone. The results show that Cr is suitable coating material for the cladding with a thin coating while Fe-Cr-Al have a better fracture mechanical performance in the cladding with thick coating. The Si-coated and SiC-coated claddings are suitable for reactors with low power fuel elements. Conclusions in this paper can provide reference and guidance for the cladding design of nuclear fuel elements.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2000.06a
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• pp.269-275
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• 2000

The fretting wear characteristics of TiN coated Zircaloy-4 tube were investigated experimentally The fretting wear experiment was performed using TiN coated Zircaloy-4 tube as the fuel rod cladding material and uncoated Zircaloy-4 tube as one of grids. TiN coating is probably one of the most frequently and successfully used PVD coatings for the mitigation of fretting wear. In this study, TiN coating by PVD was employed for improvement of Zircaloy-4 tube fretting characteristics. The fretting tester was designed and manufactured for this experiment. TiN coated Zircaloy-4 tube was used as the moving specimen, uncoated ZircaBoy-4 tube as the stationary one. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that the wear volume of TiN coated Zircaloy-4 tube increased as number of cycles, normal load and slip amplitude increase but the quantity of volume was lower than the case of uncoated Zircaloy-4 tube pairs.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.22-33
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• 1987

The confirmed integrity of nuclear fuel cladding materials is an important object during steady state and transient operations at nuclear power plant. In this context, the clad material yielding behavior is especially important because of pellet-clad gap expansion. During the steep power excursion, the in-pile irradiation behavior differences between uranium-dioxide fuel pellet and zircaloy clad induce the contact pressure between them. If this pressure reaches the zircaloy clad yield pressure, the zircaloy clad will be plastically deformed. After the reactor power resumed to normal state, this plastic permanent expansion of clad tube give rise to the pellet-clad gap expansion. In this paper, the simple mandrel expansion test method which utilizes thermal expansion difference between copper mandrel and zircaloy tube was adopted to simulate this phenomenon. That is, copper mandrel which has approximately three times of thermal expansion coefficient of zircaloy-4 (PWR fuel cladding material) were used in this experiment at the temperature range from 400C to 700C. The measured plastic expansion of zircaloy outer radius and derived mathematical relations give the yield pressure, yield stress of zircaloy-4 clad at the various clad wall temperatures, the activation energy of zircaloy tube yielding, and pellet-clad gap expansion. The obtained results are in good agreement with previous experimental results. The mathematical analysis and simple test method prove to be a reliable and simple technique to assess the yielding behavior and gap expansion measurement between zircaloy-4 tube and uranium-dioxide fuel pellet under biaxial stress conditions.

• 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.

• Journal of the Korean institute of surface engineering
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• v.35 no.5
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• pp.312-318
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• 2002

A study on the change of mechanical properties and oxidation behavior of Zircaloy-4 fuel cladding after exposing at 90$0^{\circ}C$ and $1000^{\circ}C$ for various periods of exposure time under the steam atmosphere was carried out. The growth of the $ZrO_2$ layer combined with an oxygen-rich-phase layer into the Zircaloy tube material can be described by an expression, E = 1.1√Dt + $2 $\times$ 10^{-4}$ . The tensile strength of Zircaloy tubes increased for a short period of exposure time and decreased rapidly with further exposure while the hoop strength was not decreased greatly. In the meantime, the axial and circumferential elongations of oxidized Zircaloy tubes were decreased drastically with increasing exposure time as a result of embrittlement phenomena.

• 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.

• Journal of Welding and Joining
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• v.17 no.4
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• pp.26-31
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• 1999

An attempt was made in this study to investigate the brazing characteristics of Zr-Be binary amorphous alloys for the development of a new brazing filler metal for joining Zircaloy-4 nuclear fuel cladding tubes. This study was also aimed at the feasibility study of rapidly solidified amorphous alloys to substitute the conventional physical vapor-deposited(PVD) metallic beryllium. The $Zr_{1-x}Be_{x}$($0.3\leq$x$\leq0.5$) binary amorphous alloys were produced in the ribbon form by the melt-spinning method. It was confirmed by x-ray diffraction that the ribbons were amorphous. The amorphous. the amorphous alloys were used to join bearing pads on Zircaloy-4 nuclear fuel cladding tubes. Using Zr-Be amorphous alloys as filler metals, it was found that the reduction in the tube wall thickness caused by erosion was prevented. Especially, in the case of using $Zr_{0.65}Be_{0.35}$ and $Zr_{0.7}Be_{0.3}$ amorphousalloys, the smooth and spherical primary $\alpha$-Zr particles appeared in the brazed layer, which was the most desirable microstructure from the corrosion-resistance standpoint.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.193-202
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• 2010

In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2312-2322
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• 2021

In recent studies, the creep rate of Zircaloy-4, one of the basic property parameters of the nuclear fuel code, has been commonly used with the axial creep model proposed by Rosinger et al. However, in order to calculate the circumferential deformation of the fuel cladding, there is a limitation that a difference occurs depending on the anisotropic coefficients used in deriving the circumferential creep equation by using the axial creep equation. Therefore, in this study, the existing axial creep law and the derived circumferential creep results were analyzed through a circumferential creep test by the internal pressurization method in the isothermal conditions. The circumferential creep deformation was measured through the optical image analysis method, and the results of the experiment were investigated through constructed IDECA (In-situ DEformation Calculation Algorithm based on creep) code. First, preliminary tests were performed in the isotropic β-phase. Subsequently in the anisotropic α-phase, the correlations obtained from a series of circumferential creep tests were compared with the axial creep equation, and optimized anisotropic coefficients were proposed based on the performed circumferential creep results. Finally, the IDECA prediction results using optimized anisotropic coefficients based on creep tests were validated through tube burst tests in transient conditions.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1999.06a
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• pp.303-309
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• 1999

The fretting wear behaviour of Zircaloy-4 tube used as the fuel rod cladding in PWR nuclear power plants has been investigated at the different test environment, in light water and in air as a function of slip amplitude, normal load, test duration and frequency. Zircaloy-4 tubes were used for both of oscillating and stationary specimens. A fretting wear tester was designed to be suitable for this fretting test. The wear volume and specific wear rate of Zircaloy-4 tube in water were greater than those in air under various slip amplitude. It was found that delaminate debris and surface cracks were observed at low slip amplitude and high load in water Experimental results showed that the light water accelerated the wear of Zircaloy-4 tube at low slip amplitude in fretting.