• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.255-255
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• 2002

• 한국재료학회지
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• 제12권4호
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• pp.259-263
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• 2002

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

• 열처리공학회지
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• 제17권3호
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• pp.151-164
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• 2004

The changes of microstructural and mechanical properties were evaluated for the high burn-up fuel claddings after the neutron irradiation of $1.8{\sim}3.1{\times}10^{20}n/cm^2$ (E>1.0 MEV) in HANARO research reactor. After the irradiation, the spot-type dislocations (a-type dislocations) were easily observed in most claddings, and the density of the dislocations was different depending on the grains and was higher at grain boundaries than within grains. As the final annealing temperature increased, the density of spot-type dislocations increased and the line-type dislocations (c-type dislocations) which was perpendicular to the <0002> direction, appeared sporadically in some claddings. However, the types of precipitates in the fuel claddings after the irradiation were not changed from that in unirradiated claddings. The mechanical properties including the hardness, strength and elongation after the irradiation were changed due to the formation of spot-type dislocations. That is, the increase in hardness and strength as well as the decrease in elongation after the irradiation was occurred simultaneously with increasing the final annealing temperature. Owing to the Nb contribution to the formation of spot-type dislocation during the irradiation, the increase in hardness and strength in higher Nb-contained Zr alloys after the irradiation was higher than that in lower Nb-contained Zr alloys.

• 열처리공학회지
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• 제17권4호
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• pp.223-229
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• 2004

The effects of the cooling and annealing conditions on the microstructures and corrosion properties were investigated for the Cu-contained Zr-Nb alloy (Zr-1.1Nb-0.07Cu). After annealing at $1050^{\circ}C$ for 15 min, the specimens were cooled by three methods of water quenching, air cooling, and furnace cooling. Widmanstatten structures were developed in both air- and furnace-cooled specimens, and the Widmanstatten plate width of the furnace-cooled specimens was wider than that of the air-cooled ones. The weight gain in the furnace-cooling case was higher than that in the air-cooling case. This could be the reason why the diffusion time was more enough during the furnace cooling than the air cooling. The oxide of the furnace-cooled specimen was nonunformly formed just beneath the Widmanstatten plate boundaries, where ${\beta}_{Zr}$ phases were exised concentrately. Compared with the $640^{\circ}C$ annealing after the water quenching, the $570^{\circ}C$ annealing could make the ${\beta}_{Nb}$ phases and a concomitant reduction of the Nb in the matrix, and then it could improve the corrosion resistance with the increase of the annealing time. It would be concluded that the corrosion resistance of the Zr-1.1Nb-0.07Cu was good when the Nb concentration in the matrix was reached at an equilibrium level and then the ${\beta}_{Nb}$ phase was formed.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2054-2063
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• 2020

The high-temperature oxidation behavior of ZrSi₂ used as a coating material for nuclear fuel cladding was investigated for developing accident-tolerant fuel cladding of light water reactors. Bulk ZrSi₂ samples were prepared by spark plasma sintering. In situ X-ray diffraction was conducted in air at 900, 1000, and 1100 ℃ for 20 h. The microstructures of the samples before and after oxidation were examined by scanning electron microscopy and transmission electron microscopy. The results showed that the oxide layer of zirconium silicide exhibited a layer-by-layer structure of crystalline ZrO₂ and amorphous SiO₂, and the high-temperature oxidation resistance was superior to that of Zircaloy-4 owing to the SiO₂ layer formed. ZrSi₂ was coated on the Zircaloy-4 tube surface using laser 3D printing, and the coated tube was oxidized for 2000 s at 1200 ℃ under a vapor/argon mixture atmosphere. The outer surface of the coated tube was hardly oxidized (10-30 ㎛), while the inner surface of the uncoated tube was significantly oxidized to approximately 300 ㎛.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• 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.

• 한국재료학회지
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• 제11권6호
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• pp.453-459
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• 2001

미세조직에 따라 기계적 성질 및 내산화성.내부식성 등의 제반 성질이 크게 변하는 Zr계 핵연료 피복관은 미세조직의 최적화가 중요하다. 이러한 미세조직은 합금원소의 고용도에 크게 의존하지만, Zr은 대부분의 용질합금원소의 고용도가 매우 작아서 측정이 곤란하였다. 본 연구에서는 핵연료 피복관 재료의 주요한 기본조성 재료인 Zr-0.8Sn 합금에 대한 Nb의 고용도를 TEP 측정방법을 이용하여 연구하였으며, 광학현미경과 전자현미경으로 미세조직을 관찰하여 이를 확인하였다. 균질화 처리온도가 증가함에 따라 고용된 Nb 함량이 증가하여 Zr-0.8Sn 합금의 TEP는 감소하는 경향을 보였다. 처리온도가 더욱 증가하면 TEP의 포화가 발생하였는데 이는 TEP에 영향을 미치는 고용된 합금원소의 함량 변화가 없기 때문이다. 따라서, TEP의 포화영역이 나타나기 시작하는 균질화 처리온도가 첨가된 Nb이 Zr-0.8Sn 합금에 모두 고용되는 시점이며, 이를 토대로 온도에 따른 Zr-0.8Sn 합금에서의 Nb 고용도 ($C_{Nb}$ )를 $4.69097{\times}10^{16}{\times}e^{-25300\times\;I/T}$(ppm.at.%)로 나타낼 수 있었다.

• 한국재료학회지
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• 제9권5호
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• pp.452-459
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• 1999

고연소도 핵연료피복관용 신합금 재료를 개발하기 위한 연구로 Zr-0.8Sn-xNb(x = 0.2,0.4,0.8, 1.0) 계 합금을 제조하여 Nb 첨가량이 Zr 합금의 미세구조 및 부식특성에 미치는 영향을 조사하였다. 미세조직 관찰결과 Nb첨가량이 증가함에 따라 결정립의 크기는 강소하였고 석출물의 량은 증가하였다. $360^{\circ}C$ 물 분위기에서 부식시험 한 결과 Nb 함량이 적을수록 부식저항성이 증가하는 경향을 나타냈으며며, Zr-0.8Sn-0.2Nb 합금이 가장 우수한 부식저항성을 보였다. 동얼 두께의 산화막에 대하여 XRD 분석한 결과, 내식성이 우수한 0.2 wt.% Nb 합금에서는 산화막내 tetra-$ZrO_2$의 분율이 높게 관찰되었다. 합금설계 관점에서 Zr-O.8Sn-xNb 합금계에 Nb올 첨가할때 Nb은 고용도 이하로 첨가되어야 한다고 사료된다.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3535-3542
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• 2023

Eutectic reactions of five kinds of Cr-coated Zr alloy cladding with different base materials (Zr-Nb-Sn alloy or Zr-Nb alloy), different coating thicknesses (6~22.5 mm), and different coating materials (Cr single layer or Cr/CrN bilayer) were studied using Differential Scanning Calorimetry (DSC). The DSC experiments demonstrated that the onset temperatures of the Cr single layer coated specimens were almost identical to ~1308 ℃, regardless of base materials or coating thicknesses. This study demonstrated that the Cr/CrN bilayer coated Zr-Nb-Sn alloy has a slightly (~10 ℃) higher eutectic onset temperature compared to the single Cr-coated specimen. The eutectic region characterized by post-eutectic microstructure proportionally increases with coating thickness. The post-eutectic characterization with different holding times at high temperature (1310-1330 ℃) reveals that progression of Zr-Cr eutectic requires time, and it dramatically changed with exposure time and temperature. The practical value of the time gain in non-instantaneous eutectic formation in terms of safety margin, however, seems to be limited.