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Path: Home » Indice Pubblicazioni » Convegni ATI - Accesso riservato soci CTI » CA - 65 - Cagliari 2010 » Fast running models of molten corium ...

Convegni ATI - Accesso riservato soci CTI

Fast running models of molten corium coolability for safety analysis of nuclear reactors

Pubblicazione


Autore: Parozzi F., Polidoro F., Naviglio A., Zardo G.

Collana: CA - 65 - Cagliari 2010

Note:
A design objective of nuclear plants is to greatly limit the external consequences of accidents. To reach this target in case of core meltdown, there is the need to cool the corium debris and avoid its attack to the remaining safety barriers.
The corium debris cooling could permit the direct confinement of radioactive products within the vessel. In the case of vessel failure, the corium spreading on suitable structures or core-catchers would allow a safe confinement.
The acceptance of these accident recovery strategies requires simulation models and experimental verifications, as any confinement structure must be designed for optimizing the corium cooling function (configuration, choice of materials and coolant), removing the dependence of recovery success on uncertainties related to poorly known phenomena.
In order to analyze the thermal behavior of such corium-structure-coolant systems from an engineering point of view, an original simulation tool was developed by Italian researchers in the framework of international cooperations. This tool, called
CORIUM-2D, is based on a fast-running approach characterized by high numerical stability that implies only mass and energy balances. The model calculates dynamically, and in two-dimensions, the corium and structure temperature field taking into account the heat transferred by means of: solid to solid conduction, radiation, liquid corium convection and contact with a coolant. Phase changes in structural and corium materials are also considered, then taking into account the possible formation or melting of corium crusts. The paper reports some demonstrative applications of this tool to new LWR designs, together with a preliminary code validation through theoretical benchmarks. The extension of its models to Generation IV fast reactors cooled with liquid metal is then illustrated, and an example of its application to accident scenarios hypothesized for sodium reactors is also provided.


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