CANDU-600 Heat Transport System Flow Stability

The thermohydraulic stability of the Canada deuterium uranium (CANDU)-600 heat transport system was investigated from a theoretical, numerical, and experimental point of view. Simple theoretical models, used to provide phenomenological insight as a guide to the numerical and experimental studies, showed that a major form of positive feedback existed through an interplay of circuit flow, outlet header void fraction, and outlet header pressure. The flow and pressure dynamics proved to be good indicators of system stability. System computer codes (SOPHT, FIREBIRD, and HYDNA) were used for the detailed modeling of system dynamics. These codes showed that neither Ledinegg nor parallel channel instabilities occur in CANDU-600 nuclear reactors. Loop stability was predicted under all conditions with the reactor outlet header interconnect line in service as designed. With the interconnect line disconnected, loop instability was predicted for a narrow outlet header quality range (1 to 8%). These predictions were fully confirmed by semiscale experimental loop tests and plant commissioning tests.