engleski

ASTEC Computer Code Application to NPP Krško

The ASTEC computer code was used to simulate a severe accident sequence in the nuclear power plant Krško. The main focus was on the in-vessel phase of the accident. The model of the plant included the detailed nodalization of primary and secondary circuits without the containment. The core, the reactor coolant system, steam generators, steam lines, feedwater and auxiliary feedwater pipes were modelled as a set of thermal hydraulic volumes connected by junctions, to which heat structures were attached to simulate heat losses to the environment. The ICARE module was used to model the reactor core and the CESAR module to model all the other plant systems: primary and secondary circuit piping, the pressurizer and the steam generators. Pumps, accumulators, pressurizer heaters and spray, safety and relief valves were modelled by the SYSINT module. The analysis was conducted in two steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout (SBO) accident with the leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. The analyzed SBO scenario included the loss of both off-site and on-site AC power. The only systems available were passive systems: the accumulators and the turbine driven auxiliary feedwater (AFW) system. Two scenarios were analyzed: one with and one without the AFW. Since the operation of the turbine driven AFW pump was not completely passive because it required operator action to manually control the steam generator downcomer water level, the possibility of the unavailability of any operator actions was taken into account by calculating the sequence without heat removal by the secondary system. The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with the core melt and the reactor pressure vessel failure with significant release of hydrogen. At the end, the results of the ASTEC calculation were compared with the results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results showed a good agreement between predictions of the two codes.

ASTEC; NPP; SBO; severe accident

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