In the event of a severe accident in a pressurized water reactor, the core of a reactor melts and forms corium, a mixture that includes molten UO2 and ZrO2. If the reactor pressure vessel fails, corium can be relocated in the containment cavity and interact with concrete forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question is what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using a new computer code MOCO. The code considers the heat transfer behavior in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, and incorporates phenomenology that is deemed to be important for analyzing debris cooling behavior. The interaction between thermalhydraulics and physic-chemistry is modeled in MOCO. The main purpose of this paper is to present the modeling used in MOCO and some validation calculations using the data of experiments available in the literature.
- Nuclear Engineering Division
Simulation of Molten Corium Concrete Interaction With the MOCO Code
Lin, B, Qiu, S, Su, G, Tian, W, & Zhang, Y. "Simulation of Molten Corium Concrete Interaction With the MOCO Code." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security. Prague, Czech Republic. July 7–11, 2014. V003T06A012. ASME. https://doi.org/10.1115/ICONE22-30345
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