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Past and Future Research at IRSN on Corium Progression and Related Mitigation Strategies in a Severe Accident

Didier Jacquemain, Didier Vola, Renaud Meignen, Jean-Michel Bonnet, Florian Fichot, Emmanuel Raimond, Marc Barrachin

Nuclear Technology / Volume 196 / Number 2 / November 2016 / Pages 161-174

Technical Paper /

First Online Publication:October 3, 2016
Updated:November 2, 2016

Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Sûreté Nucléaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core–concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors.

This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions.

In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis.

This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management.