Abstract
Three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal–hydraulic system codes. In the EU Phare project SRR1/95 these codes have been validated against real plant transients by the participants from several countries. Data measured during a test in the Balakovo-4 VVER-1000 have been analysed by coupled codes. In the test, one of two working feed water pumps of the steam generators was switched off at nominal power. The steady-state assembly powers measured before and after this transient are reproduced by the codes with a maximum deviation of about 5%. The time behaviour of the most safety-relevant parameters, such as total fission power, coolant temperatures and pressures is well modelled. Thermal–hydraulic feedback effects observed in the measurement are described by the codes in a consistent manner. The analyses have shown, that an accurate treatment of the heat transfer from the fuel rods to the coolant is important. In all, the results have increased the confidence in the coupled code analyses of VVER-1000 transients.
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