Naslov
NPP Krško Containment Modelling and Calculation with the ASTEC Code

Autori
Šadek, Siniša

Izvornik
FER-ZVNE/SA/DA-TR01/14-0

Vrsta, podvrsta
Ostale vrste radova, izvještaj

Godina
2014

Ključne riječi
ASTEC; NPP Krsko; containment; severe accident

Sažetak
ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krško, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls, the containment foundation, internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment TH behaviour ; however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. Aerosol and fission product behaviour in the containment were not modelled. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant containment parameters, such as compartments pressures and temperatures, is going to be discussed in the report.

Izvorni jezik
Engleski

Znanstvena područja
Elektrotehnika

POVEZANOST RADA