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首页> 美国政府科技报告 >PWR (Pressurized-Water Reactor) Pressure Vessel Integrity during Overcooling Accidents: A Parametric Analysis.
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PWR (Pressurized-Water Reactor) Pressure Vessel Integrity during Overcooling Accidents: A Parametric Analysis.

机译：过冷却事故中的压水容器pWR（压水反应堆）：参数分析。

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摘要

There are certain hypothetical accidents associated with pressurized-water reactors that can result in severe thermal shock to the reactor pressure vessel at a time when the primary-system pressure is substantial. These overcooling accidents, coupled with a reduction in fracture toughness due to the exposure of the vessel to fast neutrons, introduce the possibility of propagation of preexistent flaws on the inner surface of the vessel. In order to evaluate the magnitude of the problem and to provide a 'handbook' assessment capability, a fracture-mechanics parametric-type study was conducted for a number of postulated transients, assuming an initial flaw in the form of a long axial crack. In addition to the large-break loss-of-coolant accident, the postulated transients consisted of a constant pressure and an exponential decay of the temperature of the coolant in the downcomer. Parameters that were varied in the study included the thermal decay constant, coolant asymptotic temperature, primary system pressure, initial toughness of the material, copper concentration in the material, and crack depth. Results of the study are summarized.

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