Dissolution of uranium dioxide by molten zircaloy-4용융 지르칼로이-4에 의한 이산화우라늄의 융해

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Experiments with use of $UO_2$ crucibles containing molten Zircaloy are conducted in the range of reaction temperature from 2223 to 2373K and mole ratios of the $UO_2$/Zircaloy from 7 to 18.2 to estimate the amount of $UO_2$ pellet dissolution in molten Zircaloy cladding during severe fuel damage accidents for three different reactor type fuels. The uranium concentration in the Zircaloy melt rapidly increased during the initial few minutes of reaction time and approached a saturated value, depending on reaction temperature and the $UO_2$/Zircaloy mole ratio. Kinetics of uranium content increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The uranium solubility in the Zircaloy melt increased with increasing reaction temperature and decreasing the mole ratio of the $UO_2$/Zircaloy. These are attributed to progressive contraction of the two-phase [$(U,Zr)O_{2-x}$ + Liquid] region in the U-Zr-O ternary phase diagram and to decreasing oxygen solubility in the melt. An empirical correlation of uranium solubility in the Zircaloy melt was obtained as a function of the $UO_2$/Zircaloy mole ratio and reaction temperature. Uranium solubility in the Zircaloy melt determined experimentally also agreed with those predicted from the U-Zr-O ternary phase diagram at 2273K without use of an adjustable parameter. This empirical correlation could account well for the discrepancy of uranium solubility in the Zircaloy melt previously reported by other investigators. The amounts of uranium solubility decrease in initial oxygen-saturated Zircaloy melt for the mole ratios of the $UO_2$/Zircaloy of typical reactor type fuel geometries appeared to be indirectly estimated by experiments with use of appropriate larger mole ratio of the $UO_2$/oxygen-free Zircaloy. The fractional volumes of the $UO_2$ pellet dissolved in the initial oxygen-free and oxygen-satura...
Advisors
Yoon, Young-Ku윤용구
Description
한국과학기술원 : 원자력공학과,
Publisher
한국과학기술원
Issue Date
1994
Identifier
68967/325007 / 000865128
Language
eng
Description

학위논문(박사) - 한국과학기술원 : 원자력공학과, 1994.2, [ xii, 110 p. ]

URI
http://hdl.handle.net/10203/48825
Link
http://library.kaist.ac.kr/search/detail/view.do?bibCtrlNo=68967&flag=dissertation
Appears in Collection
NE-Theses_Ph.D.(박사논문)
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