• Title/Summary/Keyword: 우라늄합금

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Direct Determination of Molybdenum in Simulated Nuclear Spent Fuels by Inductively Coupled Plasma Atomic Emission Spectrometry (유도결합플라스마 원자방출분광법을 이용한 모의 사용후핵연료 중 몰리브덴 분석)

  • Choi, Kwang Soon;Lee, Chang Heon;Park, Soon Dal;Park, Yang Soon;Joe, Kih Soo
    • Analytical Science and Technology
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    • v.13 no.3
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    • pp.291-296
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    • 2000
  • The SIMFUEL which composition is similar to PWR nuclear spent fuels was dissolved with a acid digestion bomb. An analytical conditions of ICP-AES for the direct determination of molybdenum in the uranium matrices without separation process were investigated. Based on the effect of uranium on molybdenum intensity. the most optimum wavelengths of molybdenum were found to be 202.030 and 203.844 nm. However, the method of standard additions is applied to overcome the effects of changing background caused by analyzing the sample solutions containing high concentration of uranium and the standard calibration solutions. The relative error of two methods, direct and indirect measurements with cation exchange resin separation procedures, was less than 5%. Therefore it was possible for this procedure to directly measure molybdenum in uranium matrices without separation. And this method was also applied to the determination of several percent of molybdenum in a U-Mo alloy.

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수직형 카드뮴 증류장치의 제작 및 시운전

  • Kim, Ji-Yong;Gwon, Sang-Un;Kim, Gwang-Rak;Baek, Seung-U;Sim, Jun-Bo;Kim, Si-Hyeong;Jeong, Yong-Ju;An, Do-Hui
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2009.11a
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    • pp.331-332
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    • 2009
  • 본 연구에서는 공학규모 카드뮴 증류장치를 개발하기 위하여 실험실 규모의 수직형 카드뮴 증류장치를 제작하고, 이를 시운전하였다 공탑 실험에서는 증류탑에 온도분포를 잘 조절할 수 있었으며, 응축조의 냉각성능의 보완이 필요하였다. 이 장치를 이용하여 카드뮴 금속의 증류실험을 수행하였으며, 향후 차드뮴-우라늄 합금의 증류거동을 조사할 계획이다.

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X-Ray Tomography Based Simulation Feasibility Analysis of Nuclear Fuel Pellets (핵연료 펠릿의 X-선 단층촬영 기반 시뮬레이션 타당성 해석)

  • Kim, Jae-Joon
    • Journal of the Korean Society for Nondestructive Testing
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    • v.30 no.4
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    • pp.324-329
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    • 2010
  • Fuel rods using in nuclear power plants consist of uranium dioxide pellets enclosed in zirconium alloy(zircaloy) tubes. It is vitally important for the pellet surface to remain free from pits, cracks and chipping defects after it is loaded into the tubes to prevent local hot spots during reactor operation. This paper investigates the feasibility study for detecting surface flaws of pellets contained within nuclear fuel rod through X-ray tomography simulation. Reconstructed images used by parallel and fan-beam filtered back projection method were presented and confirmed the accessibility between simulation data and MPS(missing pellet surface) image data.

The measurement of oxygen and metal ratio of simulated spent fuels by wet and dry chemical analysis (습식 및 건식법에 의한 모의 사용후핵연료의 O/M비 측정)

  • Choi, Ke Chon;Lee, Chang Heon;Kim, Won Ho
    • Analytical Science and Technology
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    • v.16 no.2
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    • pp.117-124
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    • 2003
  • Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.