• Nuclear Engineering and Technology
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• 제21권4호
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• pp.341-345
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• 1989

Zircaloy-4 피복관의 조사손상에 영향을 줄 수 있는 부레이징 계면의 베릴륨 분포거동을 EPMA의 X-선 선분석으로 조사하여 다음과 같은 결과를 얻었다. 1) 부레이징에 의해 생성된 합금상의 베릴륨 함유량은 ~6.3mass%이었다. 2) 베릴륨의 기지금속(피복관 및 베어링 패드)내 확산은 부레이징 계면으로부터 ~5$\mu$m범위 외에서는 무시될 수 있었다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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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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• 제20권6호
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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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• 제20권6호
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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.

• 열처리공학회지
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• 제18권2호
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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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• 제2권6호
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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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• 제25권4호
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• pp.570-587
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• 1993

최근 전 세계적으로 내부2차수소 침투에 의한 것으로 보이는 핵연료의 심각한 파손이 잇달아 보고되었다. 본 논문에서는 결함 핵연료에서의 내부 수소 침투 현상이 개괄적으로 고찰된다. 먼저 핵연료의 피복관으로 사용되는 질코늄 합금의 개발사와 그 질코늄 합금이 사용되는 원자로 내의 운전조건이 소개되고 산화 질코늄 막에서의 수소의 투과성, 질코늄과 질코늄 합금에서의 수소의 최종 용해도와 침전도등 질코늄의 수소 침투에 관련된 기본 사항과 수소 침투가 기계적 강도에 미치는 악 영향등이 고찰된다. 결함 핵연료봉 내부에서 발생되는 수소의 대량 내부 침투의 메카니즘이 관찰된 제 현상을 중심으로 정성적으로 설명되고 이러한 수소의 대량 침투에 의한 핵연료치 파손심화를 줄이기 위한 노력의 일환으로 제시된 정량적 모델이 간단히 언급되고 이러한 정량적인 모델의 심도있는 개발을 위해 필요한 자료와 추후의 연구 내용이 설명된다.

• 한국표면공학회지
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• 제35권6호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.231.2-231.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.278-278
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• 1999