The present paper deals with analysis of an ex-vessel steam explosion and its structural dynamic response to confirm the integrity of a concrete structure simulating the reactor cavity compartment of the four loop PWR plant. The VESUVIUS code, which has been developed in the IMPACT project of NUPEC, was used for analysis of the molten-core coolant interaction phenomenon. The steam explosion model in VESUVIUS code contains break-up of melted jet, pre-mixing, and propagation of impulsion wave in the cavity water with a multiple-dimensional model. Initial conditions were determined based on the KROTOS steam explosion experiment performed at the JRC-Ispra in Italy. In a pre-mixing phase analysis, a melted jet entering into the water and the associated breakup phenomenon were analyzed with various water depths. The result showed that a spontaneous steam explosion was not predicted in the case of 3m water depth. However, a spontaneous steam explosion condition was onset in the case of 5m water depth. The results showed that the maximum pressure load to the cavity floor reached to 17MPa in the case of 5m water depth due to pressure propagation through fine fragmentation process. The conversion efficiency of mechanical energy was estimated to be about 2.9% in the compulsive triggering case with 5m water depth. The structural dynamic response analysis of ferroconcrete around the reactor vessel was performed by the AUTODYN-2D code, which is two-dimensional impact analysis code, applying analytical results obtained with the VESUVIUS code. The maximum value of the plasticity strain of concrete was estimated about 2.4%, and therefore calculated results indicated the possibility of local damage to the concrete structure. However, the re-bar remained in elastic region, and the integrity of a concrete structure was secured.
- Nuclear Engineering Division
Analyses of Ex-Vessel Steam Explosion and Its Structural Dynamic Response for a Typical PWR Plant
Kawabata, O. "Analyses of Ex-Vessel Steam Explosion and Its Structural Dynamic Response for a Typical PWR Plant." Proceedings of the 12th International Conference on Nuclear Engineering. 12th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 25–29, 2004. pp. 41-49. ASME. https://doi.org/10.1115/ICONE12-49105
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