• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2000년도 추계학술대회 논문집
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• pp.441-444
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• 2000

The recycling process for reuse of uranium in the spent fuels consists various unit processes and the decladding process to extract the spent fuel pellet from the zirconium-based cladding is the beginning process of the recycling. There are two methods - mechanical and chemical - in the decladding process. In this paper, the mechanical decladding device by using a motor as a driving part and a press pin to separate the pellets from tube has been developed. This device was automated and modularized to make the remote operation and maintenance easy.

• Journal of Mechanical Science and Technology
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• 제15권9호
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• 한국표면공학회지
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• 제35권5호
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• pp.312-318
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• 2002

A study on the change of mechanical properties and oxidation behavior of Zircaloy-4 fuel cladding after exposing at 90$0^{\circ}C$ and $1000^{\circ}C$ for various periods of exposure time under the steam atmosphere was carried out. The growth of the $ZrO_2$ layer combined with an oxygen-rich-phase layer into the Zircaloy tube material can be described by an expression, E = 1.1√Dt + $2 $\times$ 10^{-4}$ . The tensile strength of Zircaloy tubes increased for a short period of exposure time and decreased rapidly with further exposure while the hoop strength was not decreased greatly. In the meantime, the axial and circumferential elongations of oxidized Zircaloy tubes were decreased drastically with increasing exposure time as a result of embrittlement phenomena.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.202-211
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• 1998

Interfacial stresses at two-material interfaces and initial displacement field over the entire domain are obtained by modifying the potential energy functional with a penalty function, which enforces continuity of the stresses at the interface of two materials. Based on the initial displacement field and interfacial stresses, a new methodology to generate a continuous stress field over the entire domain has been proposed by combining the modified projection method of stress-smoothing and Loubignac's iterative method of improving the displacement field. Stress analysis is carried out on two examples made of dissimilar materials : one is a two-material cantilever composed of highly dissimilar materials and the other is a zirconium-lined cladding tube made of slightly dissimilar materials. Results of the analysis show that the proposed method provides an improved continuous stress field over the entire domain, and accurately predicts the nodal stresses at the interface, while the conventional displacement-based finite element method produces significant stress discontinuities at the interface. In addition, the total strain energy evaluated from the improved continuous stress field converges to the exact value in a few iterations.

• 센서학회지
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• 제15권4호
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• pp.297-301
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• 2006

We have fabricated a multi-mode nonlinear plastic optical fiber (POF) using organic nonlinear optical materials and demonstrated the propagation of light. The refractive indices of core and cladding are 1.5240 and 1.5172. We made a POF preform by rod-in tube method. The core diameter of the fabricated POF is about $30{\mu}m$. We evaluated the temperature characteristics of POF. The sensitivity is $0.345{\;}mW/^{\circ}C$ and the linearity of sensor was good.

• 분석과학
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• 제12권6호
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• pp.490-497
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• 1999

불활성기체용융-열전도도 측정법의 수소분석기를 개조하여 글로브박스에 설치했으며 조사핵연료피복관의 수소분석에 사용했다. Zr과 Ti 매질의 수소표준물질로 용융조제인 주석을 사용하지 않고도 수소함량 $3{\mu}g$까지 분석가능하였다. 시료의 크기가 작을수록 수소 회수율이 높았으며 지르칼로이 시료의 수소분석시 Ti 매질의 표준물질을 사용할 수 있음을 확인하였다. 수소분석에 사용한 실제 조사핵연료피복관의 평균 방사능은 10 mR/hr였으며 평균수소농도는 130 ppm이었다.

• 한국재료학회지
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• 제12권2호
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• pp.140-145
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• 2002

This study was conducted to investigate the brazing characteristics between Zircaloy-4 nuclear fuel cladding tubes and bearing pads with filler metals of amorphous $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.5) binary alloy, in which they were produced in the ribbon form by the melt-spinning metod. The crystallization behavior, stability, hardness and micro-structure of brazed zone were examined by X-ray diffraction, differential scanning calorimetry, micro-Vickers hardness test, optical microscopy, and transmission electron microscopy. $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.4) amorphous alloys were crystallized to $\alpha$-Zr with increasing the temperature, and the rest were transformed to ZrBe$_2$at higher temperatures. On the other hand, $Zr_{1-x}Be_x$(0.4$\leq$x$\leq$0.5) amorphous alloys were crystallized to $\alpha$-Zr and ZrBe$_2$, simultaneously. The thickness of the layer brazed with amorphous alloy was increased with increasing the beryllium content due to the higher diffusion of Be. The morphology of brazed layer with PVD Be filler metal showed dendrite while that brazed with amorphous alloys appeared globular. Micro-Vickers hardness of brazed zone increased as the beryllium content of filler metal was decreased.

• 소성∙가공
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• 제17권5호
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• pp.364-372
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• 2008

In this study, the circumferential creep behaviors ofpilgered advanced Zirconium alloy tubes such as Zr-Nb-O and Zr-Nb-Sn-Fe were investigated in the temperature range of $400\sim500^{\circ}C$ and in the stress range of 80$\sim$150MPa. The test results indicate that the stress exponent for the steady-state creep rate of the Zr-Nb-Sn-Fe alloy decreases with the increase of stress(from 6$\sim$7 to 4), while that of the Zr-Nb-O alloy is nearly independent of stress(5$\sim$6). The activation energy of creep deformation is found to be nearly the same as the activation energy for Zr self diffusion. This indicates that the creep deformation may be controlled by dislocation climb mechanism in Zr-Nb-O. On the other hand, the transition of stress exponent(from 6-7 to 4) in Zr-Nb Sn-Fe strongly suggests the transition of the rate controlling mechanism at high stresses. The lower stress exponent at high stresses in Zr-Nb-Sn-Fe can be explained by the dynamic deformation aging effect caused by interaction of dislocations with Sn substitutional atoms.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2010년도 춘계학술발표대회 초록집
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• pp.43-46
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• 2010

As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

• 대한기계학회논문집A
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• 제38권2호
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• pp.157-166
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• 2014

경수로 핵연료가 원자로내에서 연소되는 동안 핵연료 펠릿에서부터 피복관까지 온도해석은 핵연료 안전 해석에 있어 중요한 요소이며, 경수로 핵연료 온도 해석을 하기 위해서는 간극 모델 개발이 필수적이다. 간극 열전도도는 특성상 간극 두께값에 의존적이게 되며 이러한 특성으로 인해 다차원 간극 열전도도 모델이 비선형적 거동을 보인다. 본 연구에서는 선형화된 다차원 간극 열전도도 모델 개발을 위해 가상 연결 간극 요소를 제안하였다. 제안된 간극 연결 요소에 간극 열전도도를 적용하기 위해 등가 열전달 계수를 정의하였다. 제안된 모듈을 평가하기 위해 상용코드 ANSYS APDL 을 이용하여 열-구조 연계 해석 모듈을 구현하였으며, 다양한 예제를 통해 정확성과 수렴성을 평가하였다.