• Corrosion Science and Technology
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• v.14 no.6
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• pp.301-312
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• 2015

Corrosion of zirconium fuel cladding is known to limit the lifetime and reloading cycles of fuel in nuclear reactors. Oxide layers formed on ZIRLO4^{TM}$ cladding samples, after immersion for 300-hour and 50-day in a simulated primary water chemistry condition ($360^{\circ}C$ and 20 MPa), were analyzed by using the scanning transmission electron microscopy (STEM), in-situ transmission electron microscopy (in-situ TEM) with the focused ion beam (FIB) technique, and X-ray diffraction (XRD). Both samples (immersion for 300 hours and 50 days) revealed the presence of the ZrO sub-oxide phase at the metal/oxide interface and columnar grains developed perpendicularly to the metal/oxide interface. Voids and micro-cracks were also detected near the water/oxide interface, while relatively large lateral cracks were found just above the less advanced metal/oxide interface. Equiaxed grains were mainly observed near the water/oxide interface.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.243-249
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• 1998

상용 핵연료 피복관 재료로 사용되고 있는 Zircaloy-2와 Zircaloy-4 및 ZIRLO$^{TM}$에 대한 수소와의 반응거동 및 속도론적(kinetic) 자료를 얻기 위하여 electro-microbalance가 장착된 TGA (thermogravimetric analysis) 장치를 이용하여 30$0^{\circ}C$~50$0^{\circ}C$의 온도범위에서 1기압 수소와의 반응에 따른 무게증가를 in-situ로 측정하였다. Zircaloy와 수소와의 반응거동은 chamber내 온도상승시 생성되는 산화막에 의해 초기에는 느린 반응이 진행되는 영역이 존재하고 온도가 낮은 35$0^{\circ}C$ 이하에서는 이것이 잠복기 형태로 나타난 후 직선속도법칙(linear rate law)을 따르며 반응이 가속화되는 것으로 나타났고 40$0^{\circ}C$ 이하의 저온에서는 직선속도법칙에서 반응이 지연되는 방향으로 약간의 편차(deviation)가 관찰되었다. 그 결과 Zircaloy-2와 ZIRLO$^{TM}$가 Zircaloy-4보다 수소와의 반응속도가 빠르고 활성화에너지가 낮은 것으로 나타났으며 직선속도법칙을 근거로 하여 각각 1.1x$10^{7}$ exp(-20,800/RT)와 1.5x$10^{6}$exp(-18,000/RT) 및 6.9x$10^{7}$ exp(-23,800/RT) (mg/dm$^2$/min) 의 속도상수를 도출하였다. 또한, 열구배가 존재하지 않는 out-of-pile 조건하에서도 'sunburst' 형태의 국부적 수소침투가 발생할 수 있음이 ～l,000 ppm이상의 수소침투 시편에서 확인되었다. ～3,000ppm이상 침투하게 되면 표면에 수소화물이 농축되어 있는 hydride layer가 형성됨을 관찰하였으며 ～5,000ppm 이상의 경우에는 수소화물의 방향성이 random하였으며 특히, ZIRLO$^{TM}$ 시편의 경우에서는 원주방향으로 길게 이어진 수소화물과 기계적 성질에 치명적인 반경방향의 수소화물이 평행하게 배열된 것을 관찰하였다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.2
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• pp.231-238
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• 2003

The fuel of light water reactor is used for several years under high temperature and pressure, so it needs to be clad with high corrosion resistance material. The cladding materials must have the characteristics of low absorption of a neutron and high corrosion resistance. Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor have been used as cladding materials and Zirlo has been developed as the material for preventing the corrosion. If the fracture of the cladding tube occurs during operation, it will cause the economic loss to shut down and replace the system. So it is needed to evaluate the integrity of the cladding materials. In this paper, the tensile characteristics of the cladding materials were investigated for the basic research of fracture characteristics. Also the residual stress was analyzed to compare the tube type(original type) specimen and the flattened type specimen.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.489-494
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• 2001

The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.309-310
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• 2015

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.221-222
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.219-220
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• 2017

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.235-243
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• 2014

Surface oxidation behavior of ZIRLO (ZIRconium Low Oxidation) hulls was investigated using an X-ray photoelectron spectroscopy (XPS) technique. The effects of oxidation time (10-336 h at $500^{\circ}C$) and temperature ($400-700^{\circ}C$ for 10 h) were studied. Deconvolution results of the hulls oxidized at $500^{\circ}C$ revealed that a $ZrO_2$ phase appeared after 24 h (11.86%), and an increase in the $ZrO_2$ ratio was observed when the hulls were oxidized for 336 h (17.93%). On the other hand, the ZrO phase which employed 5.68% in the 10 h oxidized sample disappeared when the oxidation time increased to 24 h. The XPS results also showed that an increase in the oxidation temperature resulted in an increase in the ratio of ZrO, which increased from 0 to 5.68, 8.31, and 9.16% when the oxidation temperature increased from 400 to 500, 600, and $700^{\circ}C$, respectively. $ZrO_2$ phase was observed only in the sample that was oxidized at $700^{\circ}C$. The mechanism of ZrO formation was not conclusive, but it was suggested that a formation of hydroxide might have been accelerated at elevated temperatures leading to a formation of a $Zr(OH)_4$ phase. The relationship between the surface oxidation status of the hulls oxidized at $500^{\circ}C$ and their chlorination reaction feasibility was discussed, and it was suggested that the thickness of the oxide layer is an important parameter that determines the chlorination reaction feasibility.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.193-194
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• 2017

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.05a
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• pp.187-188
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• 2012