Zircaloy steam reaction and embrittlement of the oxidized zircaloy tube under postulated loss of coolant accident conditions. (Oxidation kinetics and embrittlement of zircaloy at above 1200/sup 0/C)
In a loss of coolant accident in a light water reactor, the zircaloy cladding oxidizes due to reaction with steam and becomes brittle. Research began in 1957 into this reaction of zircaloy and steam at a high temperature, and many research reports have been subsequently issued on this problem of loss of coolant accidents in particular. However, there have been great disparities in these reports concerning reaction rate constrants so that independent data are required in even basic research from the standpoint of safety. Oxidation rates and ductile changes have been studied at reaction temperature ranges of 900-1200/sup 0/C. Conversely, at temperatures above 1200/sup 0/C, the time for the sample temperature to rise to the target value cannot be ignored in comparison to the reaction time at the target value temperature, and the difficult problem in terms of the experimental technology of temperature measurement also remains. In this study, one method of solving these problems is advanced and oxidation as well as ductile changes are studied at reaction temperatures above 1200/sup 0/C. Moreover, the sample has been measured and changes in thickness of the ZrO/sub 2/ layer at reaction temperatures of 900-1200/sup 0/C and of the xi (oxide and ..cap alpha.. phase) layer have been studied. 10 refs., 13 figs., 4 tabs.
- Research Organization:
- Japan Atomic Energy Research Inst., Tokai, Ibaraki
- OSTI ID:
- 5449915
- Report Number(s):
- NUREG-tr-0014; ON: TI85901905
- Country of Publication:
- United States
- Language:
- English
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FUEL CANS
EMBRITTLEMENT
OXIDATION
LOSS OF COOLANT
THERMAL STRESSES
PWR TYPE REACTORS
REACTOR MATERIALS
ZIRCALOY 4
CHEMICAL REACTION KINETICS
DUCTILITY
MATHEMATICAL MODELS
REACTOR SAFETY
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TENSILE PROPERTIES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
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