Abstract
The effects of a rapid loss of coolant caused, for example, by the fracture of a pipe in the primary coolant circuit play an important part in the safety assessment of a water-cooled nuclear reactor. In particular, it is essential to know the rate of decrease of coolant density in the core and the rate of discharge of coolant into the containment space. It is uneconomic to obtain this information from experiments on each new reactor. Consequently computational methods, supported by experiments on basic elements of a coolant circuit, are being developed which will be applicable to a wide variety of reactors.
An experiment is described in which fundamental information on fluid flow dynamics was obtained. In the experiment, a flow of two-phase water at high pressure was established in a straight pipe 15 ft long and 0–621 in i.d. At the downstream end of the pipe a rupture disc was suddenly burst. The pressure versus time histories at three locations along the pipe were recorded, using fast response pressure transducers and a multichannel oscilloscope. Experimental results are given for pre-blowdown pressures of 500 and 1000 lb/in2 (abs.) and steam qualities of 10–30 per cent. The mass flux was 0·6 times 106 lb/ft2 h, and the flow regime was annular.
From the recorded pressure traces, the speed of the head of the rarefaction wave travelling upstream into the fluid was measured and also the pressure at critical flow conditions as functions of the pre-blowdown conditions. These measurements are compared with values calculated from steam tables.
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