The current severe accident management measure for Boiling Water Reactors (BWRs) in Japan adopts 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). This prioritizes rapid cooling of the corium and is expected to lead to more steam generation at the time of the RPV failure. 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 accident scenario has been tentatively defined as the TQUV scenario, in which the water injection is assumed to fail after the RPV depressurization. A set of MELCOR analysis models, which had been used for the Fukushima accident analyses has been adopted to represent TQUV event of BWR4/Mark-I. The effects of fuel coolant interaction (FCI) has not been considered in the current modeling. The analysis results show that in the case without pre-flooding D/W, the D/W wall temperature exceeded 200°C (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 water is continuously poured onto 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.