• Transactions of the Korean hydrogen and new energy society
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• v.15 no.4
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• pp.266-273
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• 2004

The objective of this study is an understanding of stress corrosion cracking of metals that is recognized to mostly limit the lifetime of the structural materials by comparing the features of delayed hydride cracking of zirconium alloys with those of stress corrosion cracking (SCC) of Ni-based alloys and hydrogen cracking of stainless steels. To this end, we investigated a dependence of delayed hydride cracking (DHC) velocity on the applied stress intensity factor and yield strength, and correlated a temperature dependence of the striation spacing and the DHC velocity. We reviewed a similarity of the features between the DHC of zirconium alloys, the SCC of Ni-based alloys and turbine rotor steels, and the hydrogen cracking of stainless steels and discussed the SCC phenomenon in metals with our DHC mode.

• Nuclear Engineering and Technology
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• v.43 no.5
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• pp.421-428
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• 2011

The out-of-pile mechanical performance and microstructure of recrystallized Zr-1.5 Nb-S alloy was investigated. The strength of the recrystallized Zr-1.5Nb-O-S alloys was observed to increase with the addition of sulfur over a wide temperature range, from room temperature up to $300^{\circ}C$. A yield drop and stress serrations due to dynamic strain were observed at room temperature and $300^{\circ}C$. Wavy and curved dislocations and loosely knit tangles were observed after strained to 0.07 at room temperature, suggesting that cross slip is easier. At $300^{\circ}C$, however, dislocations were observed to be straight and aligned along the slip plane, suggesting that cross slip is rather difficult. At $300^{\circ}C$, oxygen atoms are likely to exert a drag force on moving dislocations, intensifying the dynamic strain aging effect. Oxygen atoms segregated at partial dislocations of a screw dislocation with the edge component may hinder the cross slip, resulting in the rather straight dislocations distributed on the major slip planes. Recrystallized Zr-Nb-S alloys exhibited ductile fracture surfaces, supporting the beneficial effect of sulfur in zirconium alloys. Oxidation resistance in air was also found to be improved with the addition of sulfur in Zr-1.5 Nb-O alloys.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.423-424
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• 2005

The performance of HANA claddings was evaluated in out-of-pile conditions. All the performance test results revealed that HANA claddings were superior to the reference claddings such as Zircaloy-4 and A-cladding. Corrosion resistance was improved by 60 to 70% compared to the commercial claddings. Creep, burst, tensile, LOCA, wear and microstructural properties were shown to be as good as the commercial claddings.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.274.2-274
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.231-231
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• 1998

• Proceedings of the KWS Conference
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• 1994.05a
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• pp.103-105
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• 1994

• Journal of the Korean Society for Heat Treatment
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• v.17 no.5
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• pp.271-277
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• 2004

Zr-xNb alloys (x = 0.2, 0.8, 1.5 wt.%) were prepared to study the characteristics of the phase transformation in Zr-Nb system. The samples were heat treated at ${\beta}$-temperature ($1020^{\circ}C$) for 20 min and then cooled with different cooling rate. The microstructures of the specimens having the same compositions were changed with cooling rate and Nb content. The Widmanst$\ddot{a}$tten structure was observed on the furnace-cooled sample. The relationship between ${\alpha}$-Widmanst$\ddot{a}$tten and ${\beta}$-phase was the {0001}${\alpha}$//{110}${\beta}$, <11$\bar{2}$0>//<111>. The ${\beta}$-phase in Widmanst$\ddot{a}$tten structure of Zr-Nb alloys containing Nb more than solubility limit was identified as ${\beta}_{Zr}$ phase which was a stable phase at high temperature. In the water quenched samples, two kinds of martensite structures were observed depending on the Nb-concentration. The lath martensite was formed in Zr-0.2, 0.8 wt.% Nb alloys and the plate martensite having twins was formed in Zr-1.5 wt.% Nb alloy.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.425-426
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• 2005

1. The oxide thickness on the fuelled test rods was within the following range from 7 ${\mu}m$ to 17 ${\mu}m$. In general, the HANA claddings showed better corrosion behavior than the two reference alloys (A-Cladding and Zr-4). 2. The weight gains of corrosion coupons were ranged from 21 to 56 mg/$dm^2$.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.425-431
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• 2018

Different methods of axial and tangential testing and various sample geometries were investigated, and new test geometries were designed to determine the ultimate tensile strength of zirconium cladding tubes. The finite element method was used to model the tensile tests, and the results of the simulations were evaluated. Axial and tangential tensile tests were performed on as-received and machined fuel cladding tube samples of both E110 and E110G Russian zirconium alloys at room temperature to compare their ultimate tensile strengths and the different sample preparation methods.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.