Corium Interaction with Reactor Vessel
Investigation of Corium Melt Interaction with NPP Reactor Vessel Steel (METCOR)
Tech Area / Field
- FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors
8 Project completed
Senior Project Manager
Tocheny L V
Research Institute of Technology, Russia, Leningrad reg., Sosnovy Bor
- CEA / Direction de l'Energie Nucleaire (DEN), France, Saclay\nFramatome ANP GmbH, Germany, Erlangen\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe\nCEA / Institut de Radioprotection et de Surete Nucleaire, France, Saint-Paul-lez-Durance\nInstitut für Kern und Energietechnik, Germany, Karlsruhe\nForschungszentrum Rossendorf, Germany, Rossendorf\nUniversity of California / Department of Chemical and Mechanical and Environmental Engineering, USA, CA, Santa Barbara\nFortum Engineering Ltd., Finland, Vantaa
Project summaryThe ultimate goal of the proposed project is the nuclear reactor safety enhancement in case of a severe accident involving the core degradation. The subject addressed by the project is the in-depth study of physico-chemical processes taking place at core melt interaction with reactor vessel steel.
Studies in the framework of ISTC Project No 833-99 have been started in NITI on April 1, 1999 in order to experimentally evaluate currently unavailable qualitative and quantitative characteristics of core melt – vessel steel interaction. Combined thermohydrodynamic and physico-chemical processes will be considered in the study.
Due to considerable reductions in Project budgeting by ISTC (the current contract foresees funding only for the First Project phase, decision about the Second phase will be taken by the Steering Board) the Work plan was changed and scope of investigations correspondingly reduced as compared to the Project proposal.
The 1st phase carried out in the framework of ISTC Contract № 833-99, deals with determining interaction characteristics of corium with a specified composition (jointly with collaborators the following composition was chosen: 56.0 w% UO2 + 25.0 w% ZrO2 + 19 w% FeOх) with a vessel steel specimen positioned horizontally under the melt pool, both thermohydrodynamic and physicochemical processes taking place simultaneously.
The following parameters are varied for the study:
– degree of corium melt superheating;
– above-melt atmosphere, in the First phase it was either inert or air.
Additionally the First Project phase includes experimental examination of free convection in the melt pool (without interaction with steel specimen) depending on:
– melt pool depth;
– melt superheating degree.
These experiments are required for the validation of numeric code, which models the melt pool thermohydrodynamics.
The Second Project phase is proposed to include studies of high-temperature corium - vessel steel interaction depending on:
– degree of melt superheating;
– composition of above-melt atmosphere, inert and steam options are proposed for the Second phase.
In this way the Second Project phase will deal with corium melt – vessel steel specimen interaction having the following characteristics:
– corium composition 56 w% UO2 + 25 w% ZrO2 + 19 w% FeOх with steam above melt;
– corium composition 70 w% UO2 + 30 w% ZrO2 with steam and inert atmosphere above melt;
– unoxidized corium (U,Zr)O2-x in the inert above-melt atmosphere.
The implementation of Project 1st and 2nd phases will bring the following results:
1. Qualitative and quantitative characteristics of radioactive melt interaction with vessel steel obtained in comparable conditions depending on:
– corium melt superheating;
– above-melt atmosphere (air, inert, steam);
– corium composition.
2. Mechanisms of corium – vessel steel interaction.
3. Structural characteristics of corium and vessel steel samples.
The results can be used for:
- refinement of numeric models describing corium melt - vessel steel interaction processes;
- verification of calculation codes modeling free convection processes in the melt pool in terms of physicochemistry;
- calculation and safety upgrade of operated and designed reactors VVER, PWR and BWR.
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