• 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.

• Journal of Welding and Joining
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• v.15 no.6
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• pp.57-65
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• 1997

A study was undertaken to determine the influence of welding variables such as shielding and plasma gases, torch standoff, travel speed and heat input, etc. on the quality of plasma arc welds in Zircaloy-4 sheet, 2mm thick. Effect of shielding gases and their flow rates on the mechanical properties of Zircaloy-4 welds by plasma arc welding were determined in terms of tensile, bardness and bend tests. The microstructure and fracture surface of Zircaloy-4 welds were investigated by optical and scanning electron microscopies. In addition, the causes of porosity and undercut in plasma arc welds of Zircaloy-4 were also investigated. Zircaloy-4 weld bead width and depth by helium shielding gas showed a wider and deeper than those by argon. It was found that Zircaloy-4 welds with shielding gas of helium did dxhibit a little smoother and uniform weld beads than those with shielding gas of argon. It was also found that the optimum gas flow rates for Zircaloy-4 welding were 0.45l/min for plasma gas with Ar and 4.5 - 6 l/min for shielding gas with He. In addition, there was no big difference in the microstructure and fracture surface of the weld metals made by either Ar shielding gas or He shielding gas.

• Korean Journal of Materials Research
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• v.13 no.7
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• pp.473-477
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• 2003

The hydrogen redistribution induced by thermotransport at temperatures likely to be encountered in nuclear power reactors (300-$340^{\circ}C$) was investigated in modified Zircaloy-4 alloys. Modified Zircaloy-4 alloys were prepared by altering the chemical composition of Zircaloy-4; the oxygen content of Zircaloy-4 (0.1 wt%) was increased to 0.2, 0.5 and 1.0 wt%. The heat of transport ($Q^{*}$ ) for hydrogen was measured by changing the initial hydrogen and oxygen concentrations. It was found that the heat of transport was not affected by increases in the initial hydrogen concentration from 63.3 to 91.7 ppm. However, the value of $Q^{Q}$ decreased from 6.8 to 4.5 ㎉/mol as the initial oxygen concentration was increased from 0.2 to 1.0 wt%.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.277-283
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• 1996

최종 pilgering 단계에서의 가공량이 서로 다른 Zircaloy-4 피복관을 대상으로 350-50$0^{\circ}C$, 원주응력 80-150 N/$\textrm{mm}^2$의 이축응력 조건에서 크립시험이 수행되었다. Zircaloy-4 피복관의 크립변형률 및 크립변형량은 최종 pilgering 단계에서의 가공량에 비례하여 커졌다. 이를 토대로 크립모델 제시되었으며 제시된 모델은 Zircaloy-4 피복관의 크립거동을 매우 잘 모사하였다. Zircaloy-4 피복관의 크립활성화 에너지는 $\alpha$-zirconium에서의 자기확산의 활성화에너지 값과 거의 동일한 60 Lcal/mole 이므로, 크립지배기구는 전위상승이다. 따라서 가공량에 따라 크립변형률 및 크립변형률의 증대는 가공량에 따른 기지상내의 점결함의 증가 때문으로 사료된다.

• 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.

• Korean Journal of Materials Research
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• v.4 no.3
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• pp.357-363
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• 1994

The irradiation growth of the Zircaloy-4 cladding in the KOFA fuel loaded in the Kori-2 nuclear plant was measured to evaluate the irradiation growth behavior and to be compared with that of the Siemens cladding having different manufacturing process. Due to the partial recrystallization by final heat treatment, the KOFA Zircaloy-4 cladding showed a two step irradiation growth behavior such as the growth saturation and the accerlation which is typical of the fully annealed Zircaloy cladding. The difference in the measured irradiation growth rate between the KOFA and the Siemens cladding could be explained by the difference in the cladding texture which depends on the manufacturing process. From the measured irradiation growth data of Kori-2 KOFA fuel, a two-step irradiation growth model of the KOFA Zircaloy-4 cladding was derived, the accuracy of which can be more clearly verified as the measured data of the irradiation growth are accumulated in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.273-280
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• 2012

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in $500^{\circ}C$ LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

• 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.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1999.06a
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• pp.310-316
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• 1999

The fretting wear characteristics of the contact between Zircaloy-4 tube and Inconel 600 tube have investigated. Zircaloy-4 is used for fuel rod in nuclear reactor and Inconel 600 is used for tube In steam generator of nuclear power plant. A fretting wear tester was designed to be suitable for this fretting test. In this study, the number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. This study shows that the wear scar length of Zircaloy-4 and Inconel 600 increases as number of cycles, normal load and slip amplitude increase and the wear scar length of Zircaloy-4 is more longer than that of Inconel 600 due to the surface hardness.