• 제목/요약/키워드: UO2

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이중냉각 UO2 환형소결체 연소 거동

  • 방제건;양용식;김대호;구양현;서항석;권형문
    • 한국방사성폐기물학회:학술대회논문집
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    • 한국방사성폐기물학회 2011년도 추계학술논문요약집
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    • pp.231-232
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    • 2011
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Effects of $Nb_2O_5$, and Oxygen Potential on Sintering Behavior of $UO_2$ Fuel Pellets

  • Song, Kun-Woo;Kim, Keon-Sik;Kang, Ki-Won;Jung, Youn-Ho
    • Nuclear Engineering and Technology
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    • 제31권3호
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    • pp.335-343
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    • 1999
  • The effects of N $b_2$ $O_{5}$ and oxygen potential on the densification and grain growth of U $O_2$ fuel have been investigated.0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$fuel pellets were sintered at 1$700^{\circ}C$ for 4 hours in sintering atmospheres which have various ratios of $H_2O$ to $H_2$ gas. Compared with those of undoped U $O_2$ pellets, the sintered density and grain size of the 0.3 wt% N $b_2$ $O_{5}$ -doped U $O_2$ pellet increase under the $H_2O$/ $H_2$ gas ratio of 5.0$\times$10$^{-3}$ to 1.0$\times$10$^{-2}$ and under the $H_2O$/ $H_2$gas ratio of 5.0$\times$10$^{-3}$ to $1.5\times$10$^{-2}$ , respectively. The sintering of U $O_2$fuel pellets containing 0.1 wt% to 0.5 wt% N $b_2$ $O_{5}$ was carried out at 168$0^{\circ}C$ for 4 hours. The enhancing effect of N $b_2$ $O_{5}$ on the sintered density and grain size becomes larger as the N $b_2$ $O_{5}$ content increases. The solubility limit of N $b_2$ $O_{5}$ in U $O_{2}$ seems to be between 0.3 wt% and 0.5 wt%, and beyond the solubility limit the second phase whose composition corresponds near to N $b_2$U $O_{6}$ is precipitated on grain boundary. The enhancement of densification and grain growth in U $O_2$ is attributed to the increased concentration of a uranium vacancy which is formed by the interstitial N $b^{4+}$ ion in the U $O_2$ lattice.

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우라늄 및 희토류(稀土流) 산화물(酸化物)의 황화반응(黃化反應)에 대한 열역학적(熱力學的) 고찰(考察) (Study on Thermodynamic Properties of Sulfidization for Uranium and Rare Earth Oxides)

  • 이정원;이재원;강권호;박근일
    • 자원리싸이클링
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    • 제21권1호
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    • pp.66-74
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    • 2012
  • 우라늄 및 희토류(RE: rare-earth) 산화물의 황화반응에 대한 $M-O_2-S_2$ 상태도 및 Gibbs 자유에너지 변화(${\Delta}G^{\circ}$)와 같은 열역학적 특성 자료를 비교, 분석하여 우라늄 및 회토류 산화물의 혼합상에서 황화반응에 의해 희토류산화물만 희토류황화물로의 선택적 반응이 가능한지를 조사하였다. 황화제로는 $CS_2$가 적합하였는데, $CS_2$는 다른 황화제보다 강한 황화재이며 반응온도를 낮출 수 있다. $CS_2$를 황화제로 이용하여 $U_2-O_2-S_2$$RE-O_2-S_2$의 상태도를 비교한 결과, $UO_2$은 반응하여 UOS로 전환되며 희토류산화물은 반응하여 희토류황화물이 되었다. 희토류산화물의 황화반응에 의한 ${\Delta}G^{\circ}$는 우라늄산화물의 ${\Delta}G^{\circ}$보다 낮았다. 따라서 희토류와 우라늄 산화물 혼합상은 $300{\sim}800^{\circ}C$에서의 황화반응 시에 평형상태에서 우라늄산황화물과 희토류황화물이 우선적으로 생성된다.

내부 압력변화에 대한 사용후핵연료 분말화장치 가열로의 열 응력 해석 (Thermal Stress Analysis of Spent Fuel Vol-oxidizer Furnace on the Internal Pressure)

  • 김영환;정재후;홍동희;박병석;이종광;윤지섭
    • 한국소성가공학회:학술대회논문집
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    • 한국소성가공학회 2006년도 춘계학술대회 논문집
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    • pp.136-140
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    • 2006
  • We are developing a vol-oxidizer which transforms the spent $UO_2$ pellets into the $U_3O_8$ power through oxidizing process. The vol-oxidizer consists of furnace, filter, heater and valve etc. When the filter is blocked by the powder, the internal pressure of the furnace is increased owing to the air flow restriction. Then, the furnace vessel is swelled and deformed by it. In this paper, we proposed a procedure of the thermal analysis for furnace vessel design of spent fuel vol-oxidizer. In this work, we determined the thickness of the furnace through analyzing the internal pressure and the thermal stress of the furnace with respect to various pressure and temperature. To analyze the thermal stress, we used ANSYS 8.0 for constructing a FEM model of the furnace, and then analyzed it based on the ASME code. We also surveyed the material property and yield stress of SUS304 with various temperature. Analysis results are compared with the yield stress of the material. We manufactured the furnace and conduct the verification experiments.

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Improving the Neutronic Characteristics of a Boiling Water Reactor by Using Uranium Zirconium Hydride Fuel Instead of Uranium Dioxide Fuel

  • Galahom, Ahmed Abdelghafar
    • Nuclear Engineering and Technology
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    • 제48권3호
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    • pp.751-757
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    • 2016
  • The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.