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Modeling of corium spreading under water layer - validation on the large mass prototypic PLINIUS-2 platform

Abstract : Corium coolability after a postulated severe accident involving core meltdown and RPV failure is an important issue. This article deals with spreading and cooling of a corium in a water layer.Currently, the THEMA code, developed at CEA with EDF sponsorship , deals only with the spreading of corium on dry surface with a radiative-convective exchange coefficient. The spreading is then mainly controlled by inertial and viscous forces.In the presence of a water layer in the reactor pit, corium spreading is principally controlled by the yield stress in crust at the flow front. This required the development of a dedicated model. First, the corium crust formation modeling (upper and flow front) was needed. Thanks to this model, the crust thickness evolution with time can be described. An analogy with FinketGriffiths [1] model of volcanic lava was made.Parametric studies show that for a given flow rate, higher yield stresses gives higher height and smaller radius, and with the same stress, the larger the flow rate is, the smaller the height is and the larger the radius is.The aim is to improve the THEMA code taking into account the corium spreading in a water layer. For this, the following modifications should be done-The forces related to the yield stress in crust at the flow front have to be implemented in the momentum balance equation of THEMA; -The power associated with the tensile strength of the crust has to be added in the energy balance equation of THEMA.Finally, to validate the model, some experiments with large mass of prototypic corium are proposed. Indeed, as crust yield stress values and conductivity of corium crust are essential but very poorly known, first, dedicated experiments will be considered to measure these thermophysical data used in the model. Then, the validation of this new version of THEMA will require experiments in more representative conditions which could be carried out in the future PLINIUS-2 experimental platform in Cadarache.
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  • HAL Id : hal-02441951, version 1

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J. Haquet, L. Buffe. Modeling of corium spreading under water layer - validation on the large mass prototypic PLINIUS-2 platform. NUTHOS-11 - The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Oct 2016, Gyeongju, South Korea. ⟨hal-02441951⟩

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