• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.341-345
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• 1989

The distribution behaviors of beryllium which may produce a deleterious damage in the zircaloy cladding have been investigated by the X-ray line scanning of EPMA. The results obtained are as follows; 1) The alloy phase formed by the brazing contains ~6.3 mass % of beryllium. 2) The beryllium diffusion in the base metal (cladding and bearing pad) is recognized only in the range ~5 $\mu$m from the brazing interface.

• 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.

• 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.

• Journal of the Korean Society for Heat Treatment
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• v.18 no.2
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• pp.112-118
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• 2005

Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.570-587
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• 1993

Recently a number of severe fuel degradation events, seemingly due to internal secondary hydriding, have been reported. This paper reviews internal hydriding of defected zircaloy cladding. First, the history of zircaloy cladding development and the environment of the zircaloys in service in the nuclear reactor are introduced. Fundamental aspects of zircaloy hydriding, such as hydrogen permeability in zirconium oxide, terminal solubility and precipitation in zirconium and its alloys, and the deleterious effect of hydrides are reviewed. The mechanism of massive internal hydriding in defected zircaloy fuel rods is qualitatively described based on the observed phenomena. Significant factors affecting the hydriding process are discussed. A quantitative model for the massive hydriding as a part of an effort to mitigate fuel degradation is briefly mentioned and necessary information and recommended future work for improvement of the model are outlined.

• Journal of the Korean institute of surface engineering
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• v.35 no.6
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• pp.363-370
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• 2002

Empirical equations for the oxide thickness and the weight gain of Zircaloy-4 cladding during the oxidation in high temperature, high pressure steam have been developed. Firstly, the empirical equations for oxide thickness in 1 atm steam in 700~100$0^{\circ}C$ were made, then, the enhancement factor for the steam pressure effects on Zircaloy-4 cladding oxidation in high temperature steam was added. Based on the analysis of the weight fraction of dissolved oxygen in metal layer, empirical equations for the weight gain of Zircaloy-4 in high pressure, high temperature steam were developed. We compare the developed empirical equations with the Baker-Just correlation. The Baker-Just correlation can give a non-conservative estimation of oxidation of Zircaloy-4, depending on the steam pressure. These developed empirical equations can be used for the correct estimation of oxidation of Zircaloy-4 during accident analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.231.2-231.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.278-278
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• 1999