Corium formation from reactor components and how its properties affect later stages of a severe nuclear accident.

The early stages of a severe accident are characterised by a loss of cooling capability in the reactor core and subsequent rise in temperature of the fuel & cladding to well in excess of typical operating temperatures. This results in various interactions and reactions between the UO2 fuel, Zircaloy cladding, moderator and reactor internal structures to produce a high temperature aggressive liquid, or corium, capable of dissolving the remaining reactor internal materials (in-vessel scenario), or even penetrate the reactor vessel wall and react with the basemat of multiphase concrete (ex-vessel scenario). Research on these various and complex high-temperature systems is clearly needed on a broad front. This includes both the determination of single compound thermodynamic properties and large-scale integral tests to test complex corium phases. Experimental data can support and validate modelling of properties of simpler and more complex systems. At JRC-ITU small scale testing of typical high temperature systems for the in-vessel scenario, as well as examination of irradiated fuel/cladding interactions has been and is carried out in the context of the main international research programmes on severe accidents. This paper includes data on the UO2-PUO2 phase diagram and the U-Zr-Fe-O system melting ranges. These will be compared with other results of corium structure analysis stemming from major irradiated fuel examinations such as Phébus PF project or the TMI-2 post-accident investigation programmes. This comparison can help elucidate the underlying mechanisms of the key reactions occurring during a severe accident in a nuclear power plant, enable better prediction of the likely progression and outcome of the event, and also help improve the overall accuracy of the severe accident codes.

BOTTOMLEY Paul;  RONDINELLA Vincenzo;  PAPAIOANNOU Dimitrios;  BREMIER Stephan;  POEML Philipp;  MANARA Dario;  SOMERS Joseph;  LAJARGE Patrick; 

2014-12-09

American Nuclear Society

JRC89130

https://publications.jrc.ec.europa.eu/repository/handle/JRC89130,