Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction
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概要
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In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium–water reactions. If the leak exceeds an intermediate level (∼2 kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition, a model of the phenomenon is derived through a series of tests on sodium–water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure — occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep — upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure, and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators, in order to provide a safety margin, a time factor — i.e., the safety factor indicating multiple of actual time to failure — of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.
- 一般社団法人 日本原子力学会の論文
- 2004-06-25
著者
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Tanabe Hiromi
Japan Nuclear Cycle Development Institute
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HAMADA Hirotsugu
Japan Nuclear Cycle Development Institute