Naslov
Reactor Vessel Modelling with the MELCOR Code

Autori
Šadek, Siniša ; Grgić, Davor ; Vuković, Franjo ; Benčik, Vesna

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 2019 / Smodiš, Borut ; Udir, Nina - Ljubljana : Nuclear Society of Slovenia, 2019

ISBN
978-961-6207-47-8

Skup
International Conference Nuclear Energy For New Europe (NENE)

Mjesto i datum
Portorož, Slovenija, 09.09.2019. - 12.09.2019

Vrsta sudjelovanja
Predavanje

Vrsta recenzije
Međunarodna recenzija

Ključne riječi
PWR, MELCOR, Krsko, reactor vessel model

Sažetak
MELCOR is an integral severe accident code that enables calculation of complete transient sequence in a reactor coolant system (RCS), nuclear power plant (NPP) secondary side and the containment. NPP modelling with the MELCOR code provides the user freedom to develop its own modelling approach by following general guidelines determined by code developers. Thus, the nodalization is not prescribed but it is still important to create a model in a logical manner that reflects the system design and operation. The code MELCOR 1.8.6 is used for a calculation of a postulated station blackout accident in the NPP Krško. The whole sequence of events during a severe accident is covered, including the thermal-hydraulic behaviour of the RCS, core cladding oxidation, fuel elements degradation and melt-down, molten corium concrete interaction in the containment cavity, the containment heat-up and pressurization. The focus of the paper is on the reactor core behaviour and influence on timing of core damage propagation events caused by different reactor pressure vessel (RPV) and core thermal-hydraulic models. Two models of the RPV are developed: one that uses complex nodalization of the RPV upper plenum, the core and the lower plenum, and the other one that uses a coarse nodalization scheme for these components. For example, in the first NPP model, the core is modelled with 12 axial control volumes, while in the second one, it is modelled with only one control volume representing the whole core region. In both models, the core components are radially divided in five regions. The similarities and differences in the analysis results between these two approaches are presented and discussed in the paper, with special attention given to the RCS thermal-hydraulic behaviour and the core damage progression.

Izvorni jezik
Engleski

Znanstvena područja
Elektrotehnika, Strojarstvo

POVEZANOST RADA