• 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.

• 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.

• Tribology and Lubricants
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• v.15 no.1
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• pp.83-89
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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 was greater than those in air under various slip amplitude. Delaminates and surface cracks were observed at low slip amplitude and high load of fretting test in water, but the traces of adhesion and plowing were observed at and above 200 Um. The water accelerates the wear of Zircaloy-4 tube at lower 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.

• Tribology and Lubricants
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• v.14 no.4
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• pp.88-94
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• 1998

The fretting wear characteristics of Zircaloy-4 tube in light water were investigated experimentally. A fretting wear tester was designed to be suitable for this fretting test. This study was focused on the effects due to the combination of normal load, slip amplitude and number of cycles as the main factors of fretting. The results of this study showed that the wear volume increased abruptly at slip amplitude above 100 ${\mu}{\textrm}{m}$, which is defined as critical slip amplitude of Zircaloy-4 tube in light water, and that under 160 ${\mu}{\textrm}{m}$ the wear volume decreased as load increased at the same slip amplitude.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1998.04a
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• pp.55-63
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• 1998

The fretting wear characteristics of Zircaloy-4 tube in light water were investigated experimentally. A fretting wear tester was designed to be suitable for this fretting test. This study was focused on the effects due to the combination of normal load, slip amplitude and number of cycles as the main factors of fretting. The results of this study showed that the wear volume increased abruptly at slip amplitude above 100$\mu$m, which is defined as critical slip amplitude of Zircaloy-4 tube in light water, and that under 160$\mu$m the wear volume decreased as load increased at the same slip amplitude.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.256-261
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• 2001

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 to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• 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.

• 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.

• Journal of Welding and Joining
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• v.20 no.6
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• pp.796-801
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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($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.