• Analytical Science and Technology
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• v.13 no.5
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• pp.601-607
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• 2000

Lanthanum has been used as one of the burnup monitor in spent nuclear fuel. $U_3Si/Al$ spent nuclear fuel contains small amount of La in high concentration of U and Al. Therefore, chemical separation of La is required to remove matrix elements. At first, ion chromatography (IC) and inductively coupled plasma systems were installed in radiation shielded glove box to handle the radioactive samples. Retention behavior of uranium, aluminum, lanthanum and some interesting fission products (Sr, Zr, Y, Mo, Ru, Pd, Rh, Cs, Ba, Ce, Pr, Nd, Sm, Eu and Cd) was investigated using the CG10 column and ${\alpha}$-HiBA eluent. As all elements were eluted earlier than lanthanum in 0.2 M ${\alpha}$-HiBA eluent, a portion of U and Al was directly passed to waste using a three way valve between the column and the nebulizer. Thus it was possible to determine the lanthanum in a high concentration of U and Al matrix. Retention time of La was about 12 minutes in this separation condition. Optimum range for the determination of La in $U_3Si/Al$ spent nuclear fuel was $1-10{\mu}g/L$ (ppb) with this system and detection limit was $0.25{\mu}g/L$ in case of $200{\mu}L$ of sample volume.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2937-2952
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• 2021

A typical three-dimensional finite element model for a fuel assembly is established, which is composed of 16 monolithic U-10Mo fuel plates and Al alloy frame. The distribution and evolution results of temperature, displacement and stresses/strains in all the parts are numerically obtained and analyzed with a self-developed code of FUELTM. The simulation results indicate that (1) the out-of-plane displacements of Al alloy side plates are mainly attributed to the bending deformations; (2) enhanced out-of-plane displacements appear in fuel plates adjacent to the outside Al plates, which results from the occurred bending deformations due to the applied constraints of outside Al plates; (3) an intense interaction of fuel foil with the cladding occurs near the foil edge, which appears more heavily in the fuel plates adjacent to the outside Al plates. The maximum first principal stresses in the fuel foil are similar for all the fuel plates and appear near the fuel foil edge; while, the through-thickness creep strains of fuel foil in the fuel plate near the central region of fuel assembly are larger, and the induced creep damage might weaken the fuel skeleton strength and raise the fuel failure risk.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.08a
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• pp.266-267
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• 2011

질화갈륨(GaN)은 높은 전자이동도 및 높은 항복전계를 가지며 낮은 온저항으로 인하여 에너지효율이 우수하기 때문에 고출력 전력소자 분야에서 많은 관심을 받고 있다. GaN을 이용한 고출력 전력소자의 경우 상용화 수준에 근접할 만한 기술적 진보가 있었으나, 페르미 레벨 고정(Fermi-level pinning) 현상, 소자의 누설전류 등 아직 해결되어야 할 문제를 갖고 있다. 본 연구에서는 실리콘 기판 위에 성장된 GaN 에피탁시를 활용한 고출력 전력소자의 누설전류를 억제시키기 위해 오믹 접합 중 Au의 상호확산을 억제하는 중간층 금속(Mo or Ni)을 변화시켰으며 오믹 열처리 온도에 따른 특성을 비교 연구하였다. $Cl_2$와 $BCl_3$를 이용하여 0.6 ${\mu}m$ 깊이의 메사 구조가 활성영역을 형성하였고, Si 도핑된 n-GaN 위에 Ti/Al/Mo/Au (20/100/25/200 nm) 와 Ti/Al/Ni/Au (20/100/25/200 nm) 오믹 접합을 각각 설계, 제작하였다. 오믹 열처리시의 GaN 표면오염을 방지하기 위해 $SiO_2$ 희생층을 증착하였다. 오믹 접합 형성을 위해 각 750$^{\circ}C$, 800$^{\circ}C$, 850$^{\circ}C$에서 30초간 열처리를 진행 하였으며, 이후 6 : 1 BOE 용액으로 $SiO_2$ 희생층을 제거하였다. 750, 800, 850$^{\circ}C$에서 Ti/Al/Mo/Au 구조의 오믹 접합 저항은 각 2.56, 2.34, 2.22 ${\Omega}$-mm 이었으며, Ti/Al/Ni/Au 구조의 오믹 접합 저항은 각 43.72, 2.64, 1.86 ${\Omega}$-mm이었다. Isolation 누설전류를 측정하기 위해서 두 개의 오믹 접합 사이에 메사 구조가 있는 테스트 구조를 제안하였다. Isolation 누설전류는 Ti/Al/Mo/Au 구조에서 두 오믹 접합 사이의 거리가 25 ${\mu}m$이고 100 V일 때 750, 800, 850 $^{\circ}C$의 열처리 온도에서 각 1.25 nA/${\mu}m$, 2.48 nA/${\mu}m$, 8.76 nA/${\mu}m$이었으며, Ti/Al/Ni/Au 구조에서는 각 1.58 nA/${\mu}m$, 2.13 nA/${\mu}m$, 96.36 nA/${\mu}m$이었다. 열처리 온도가 증가하며 오믹 접합 저항은 감소하였으나 isolation 누설전류는 증가하였다. 750$^{\circ}C$ 열처리에서 오믹 접합 저항은Ti/Al/Mo/Au 구조가 Ti/Al/Ni/Au 구조보다 약 17배 우수하였고, 850$^{\circ}C$ 고온의 열처리 경우 Ti/Al/Mo/Au 구조의 isolation 누설전류는 8.76 nA/${\mu}m$로 Ti/Al/Ni/Au의 누설전류 96.36 nA/${\mu}m$보다 약 11배 우수하였다. Ti/Al/Mo/Au가 Ti/Al/Ni/Au 보다 오믹 접합 저항과 isolation 누설전류 측면에서 전력용 GaN 소자에 적합함을 확인하였다.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Proceedings of the Materials Research Society of Korea Conference
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• 2005.05a
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• pp.142-142
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.237.1-237
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• 1999

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1575-1588
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• 2019

Fracture near the U-10Mo/cladding material interface impacts fuel service life. In this work, a mesoscale stress model is developed with the fuel foil considered as a porous medium having gas bubbles and bearing bubble pressure and surface tension. The models for the evolution of bubble volume fraction, size and internal pressure are also obtained. For a U-10Mo/Al monolithic fuel plate under location-dependent irradiation, the finite element simulation of the thermo-mechanical coupling behavior is implemented to obtain the bubble distribution and evolution behavior together with their effects on the mesoscale stresses. The numerical simulation results indicate that higher macroscale tensile stresses appear close to the locations with the maximum increments of fuel foil thickness, which is intensively related to irradiation creep deformations. The maximum mesoscale tensile stress is more than 2 times of the macroscale one on the irradiation time of 98 days, which results from the contributions of considerable volume fraction and internal pressure of bubbles. This study lays a foundation for the fracture mechanism analysis and development of a fracture criterion for U-10Mo monolithic fuels.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.169-182
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• 2014

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.168-168
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• 2012

본 연구에서는 Molybdenum oxide (MoOx)-doped 4,4',4"-tris[2-naphthyl(amino)] triphenylamine(2-TNATA)의 P-doping에 의한 hole ohmic contact과 fullerene (C60)/lithium (LiF)의 electron ohmic contact에 의한 All Ohmic contact를 이용한 유기 발광 다이오드 (OLEDs)의 광저항 특성의 향상을 설명한다. 이 소자의 성능은 MoOx-doped 2-TNATA의 두께와 도핑농도에 큰 영향을 받는다. glass/ITO/MoOx-doped 2-TNATA (100 nm)/Al 구조의 소자에서 MoOx-doped 2-TNATA 도핑 농도가 25%에서 75%로 증가할수록 hole only device의 hole ohmic 특성이 향상됐다. 그 이유는 p-type doping effect 때문이다. 또한 photoemission spectra 분석결과, p-type doping effect는 hole-injecting barrier 높이는 낮추고, hole conductivity는 향상되었다. 이것은 2-TNATA에 도핑된 MoOx의 전하전송 콤플렉스의 형성으로 hole carrier의 수가 증가하여 발생되었다. MoOx-doped 2-TNATA의 hole ohmic contact과 fullerene (C60)/lithium fluoride (LiF)의 electron ohmic contact 으로 구성된 glass/ITO/MoOx-doped 2-TNATA (75%, 60 nm)/NPB (10 nm)/Alq3 (35 nm)/C60 (5 nm)/LiF (1 nm)/Al (150 nm)의 소자구조는 6,4V에서 127,600 cd/m2 최대 휘도와 약 1,000 cd/m2에서 4.7 lm/W의 높은 전력 효율을 보여준다.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.979-986
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• 2013

Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 $g-U/cm^3$ were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional $UAl_x$ dispersion targets, while increasing the uranium density in the target plates.