• Nuclear Engineering and Technology
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• 제18권3호
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• pp.218-227
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• 1986

가상적인 냉각제 상실 사고시의 조건하에 일어날 수 있는 취약화 현상에 대한 자료를 얻기 위하여 고온의 수중기 분위기에서 Zircaloy-4 핵연료피복관의 산화거동과 기계적성질 변화에 대한 연구를 수행하였다. 시편은 캔두형핵연료 피복관으로 사용되는 질칼로이 튜브를 사용하였으며 냉각제 상실 사고시 야기될 수 있는 수중기 분위기속 90$0^{\circ}C$와 1,00$0^{\circ}C$에서 유지시간을 변경하여 가면서 산화시켰다. 질칼로이 피복관의 표면과 내부에서 ZrO$_2$ 및 $\alpha$상의 형성속도 E는 온도와 시간의 함수인 E=1.1√Dt+0.002로 나타났다. 여기서 D는 온도에 의존하는 화산계수임. 시편에 대한 인장강도, 후프강도 및 연신율을 측정한 결과 단시간 산화된 시편의 인장강도는 원래의 피복관에 비해 처음에는 약간 증가하다가 계속되는 유지 시간에 따라 감소하였다. 후프강도는 유지 시간에 따라 많이 감소하지 않았으며 외경 방향의 인장율을 급격히 감소하였다. 피복관의 선택 방위 측정 결과 원래의 피복관 입자는 대부분이 기저면(0001)에 대한 극축이 외경 방향에 평행하게 놓였었으나 1,00$0^{\circ}C$에서 열처리한 경우는 극축이 외경 방향에 수직으로 변경됨을 알 수 있었으며 이러한 결정면의 방위분포 결과가 후프강도의 유지에 기여하는 것으로 추측되었다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2041-2046
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• 2020

The present article investigates the influence of hydrogen concentration on the creep performance of cold-worked stress-relieved unirradiated Zircaloy-4 cladding tube using nanoindentation technique. The as-received Zircaloy-4 tube is hydrided to the concentrations of 600 ppm and 900 ppm using gaseous hydrogen charging method. Constant load indentation creep tests are performed for a dwell period of 600 s in the temperature range of 300℃-500 ℃ at 1000 μN, 2000 μN, and 3000 μN. The impact of hydrogen is evaluated in terms of steady state power law creep exponent and activation energy. The power law creep exponent decreases with increase in hydrogen concentration, however, it remains fairly constant with increase in temperature up to 500 ℃. Moreover, activation energy too decreases significantly with increase in hydrogen concentration. The mean stress exponent and activation energy are found to be 3.58 and 28.67 kJ/mol, respectively, for as-received sample.

• Tribology and Lubricants
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• 제17권6호
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• pp.453-458
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• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• Tribology and Lubricants
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• 제25권4호
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• pp.250-255
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• 2009

Oxidation characteristics of Zirlo and Zircaloy-4 tubes, which are widely used as nuclear power fuel cladding, are studied in steam environment up to $1200^{\circ}C$. Oxidation resistances are compared in terms of the mass increase due to the absorption of oxygen. The evolution of microscopic structure accompanied with the oxidation process is investigated. Also, the influence of oxidation on the fretting wear characteristics of the tubes is studied. Piezo-electrically actuated rig is employed to fret the tubes with cross-contacting arrangement. Wear scar is observed and measured, by using microscopes and a 3D-profiler. The results of fretting wear are quantified in terms of scar size, wear volume and wear coefficient, and compared for the three different tube materials of oxidated Zirlo, virgin Zirlo and Zircaloy-4.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• 한국재료학회지
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• 제14권7호
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• pp.451-456
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• 2004

Thermal creep properties of the zirconium tube which was developed for high burnup application were evaluated. The creep test of cladding tubes after various final heat treatment was carried out by the internal pressurization method in the temperature range from $350^{\circ}C to 400^{\circ}C$ and from 100 to 150 MPa in the hoop stress. Creep tests were lasted up to 900days, which showed the steady-state secondary creep rate. The creep resistance of zirconium claddings was higher than that of Zircaloy-4. Factors that affect creep resistance, such as final annealing temperature, applied stress and alloying element were discussed. Tin as an alloying element was more effective than niobium due to solute hardening effect of tin. In case of advanced claddings, the optimization of final heat treatment temperature as well as alloying element causes a great influence on the improvement of creep resistance.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2000년도 제31회 춘계학술대회
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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.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.249-258
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• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.247.2-247
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• 2001