Pregled bibliografske jedinice broj: 670856
Severe Accident Analysis in the NPP Krško with the ASTEC Code
Severe Accident Analysis in the NPP Krško with the ASTEC Code // Proceedings of the International Conference Nuclear Energy for New Europe 2013 / Cizelj, Leon ; Leskovar, Matjaž ; Uršič, Mitja (ur.).
Ljubljana: Nuclear Society of Slovenia, 2013. str. 409.1-409.12 (poster, međunarodna recenzija, cjeloviti rad (in extenso), znanstveni)
CROSBI ID: 670856 Za ispravke kontaktirajte CROSBI podršku putem web obrasca
Naslov
Severe Accident Analysis in the NPP Krško with the ASTEC Code
Autori
Šadek, Siniša ; Amižić, Milan ; Grgić, Davor
Vrsta, podvrsta i kategorija rada
Radovi u zbornicima skupova, cjeloviti rad (in extenso), znanstveni
Izvornik
Proceedings of the International Conference Nuclear Energy for New Europe 2013
/ Cizelj, Leon ; Leskovar, Matjaž ; Uršič, Mitja - Ljubljana : Nuclear Society of Slovenia, 2013, 409.1-409.12
ISBN
978-961-6207-36-2
Skup
International Conference Nuclear Energy for New Europe 2013
Mjesto i datum
Bled, Slovenija, 09.09.2013. - 12.09.2013
Vrsta sudjelovanja
Poster
Vrsta recenzije
Međunarodna recenzija
Ključne riječi
severe accident; ASTEC; NE Krško
Sažetak
The ASTEC/V2.0 computer code was used to simulate a severe accident sequence in the nuclear power plant Krško which is a two-loop pressurized water reactor plant. ASTEC is an integral 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. 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 accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater. 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. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those two codes.
Izvorni jezik
Engleski
Znanstvena područja
Elektrotehnika
POVEZANOST RADA
Projekti:
036-0361590-1589 - Nuklearne elektrane za održivu proizvodnju električne energije (Feretić, Danilo, MZO ) ( CroRIS)
Ustanove:
Fakultet elektrotehnike i računarstva, Zagreb
