The current severe accident management measure for Boiling Water Reactors (BWRs) in Japan adopt pre-flooding of the drywell (D/W) floor when the reactor pressure vessel (RPV) is expected to fail to prevent failure of the primary containment vessel (PCV) (wet cavity strategy).&nbsp; Compared with the other possible strategy, which is to keep the D/W floor dry until the RPV failure (dry cavity strategy), the current strategy prioritizes rapid cooling of the molten core materials (corium) and is expected to lead to more steam generation at the time of the RPV failure as the hot corium interacts with the pre-flooded water pool. Consequently, PCV overheating can be well prevented, but early over pressurization of the PCV is expected. Then, D/W spray is activated to suppress the PCV pressure and venting from the W/W or D/W may be considered when the PCV pressure is expected to exceed the design pressure. Thus, although features of the above explained strategies are relatively well understood qualitatively, there seems to be lack of quantitative comparisons of the two strategies with respect to prevention of the PCV overheating, over pressurization and ultimately, suppressing release of the source term to the environment when it is inevitable. Based on these understandings, the purpose of this study is to evaluate the influences of changes in the accident management measure on PCV integrity / failure mode and the amount of radioactive material released to the environment for BWRs. The integrated severe accident analysis code MELCOR-2.2 is used. The accident scenario has been tentatively defined as the TQUV scenario, in which the water injection is assumed to fail after the RPV depressurization.A MELCOR analysis model, which had been used for the Fukushima accident analyses is been tentatively adopted to represent TQUV event of BWR4/Mark-I. The pedestal and D/W are represented by 2-D equivalent geometries, with considerations to conserve the ratios of volume to surface areas. The effects of fuel coolant interaction (FCI), which involve fragmentation of corium in water pool and the resulting rapid steam generation has not been considered in the current modeling. Basaltic concrete is assumed for the PCV floor to model molten-core concrete interaction (MCCI).Different combinations of accident measures have been considered with regard to with / without pre-flooding of the D/W floor, continuous water injection after RPV failure, D/W spray, W/W venting and D/W venting. The analysis results show that in the case without pre-flooding D/W, the D/W wall temperature exceeded 200℃ (which may be regarded as a rough estimate for the PCV failure by overheating) in about 30 minutes after the RPV failure. On the other hand, if the D/W floor is flooded in advance and water is continuously poured into the D/W floor after the RPV failure, overheating failure of the PCV can be avoided, but over-pressure failure may occur within a few hours after the RPV failure.

Analyses of Wet and Dry Cavity Strategies for Bwr Severe Accident Management With Melcor-2.2

Paper Proceedings: https://asmedigitalcollection.asme.org/ICONE/ICONE28/volume/85246

2021 28th International Conference on Nuclear Engineering (ICONE28)

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