MELCOR 2.2 iPWR LOCA type accident analysis, PART II: BDBA

The integrated Pressurized Water Reactor brings many potential improvements for the nuclear industry, such as passive and modular design, which potentially supports reliability and safety. Besides their advantages, passive systems are also more challenging to simulate, and predictions of the complex behavior of the reactor, especially under accident conditions, need to be validated. In the first part of this study, the authors analyzed the thermal-hydraulic response of the iPWR to several design basis accident sequences without entering into the severe accident domain. As a continuation of the investigation, a sensitivity study of the beyond design basis accident scenario was performed and analyzed as described and presented here, part 2 of the study.

A generic iPWR MELCOR 2.2 input deck was developed and used to perform a loss-of-coolant accident (LOCA)-type scenario analysis in which a break is assumed in the chemical and volume control system line. The effect of the elevation of the break and decay heat on accident progression is investigated. This allows the examination of input deck and code reliability under different conditions, from full core uncovery to mitigated accidents. Overall, eight cases were calculated in which the break elevation and decay heat were varied, providing knowledge about the modeling of the iPWR design and potential analytical challenges, which was the main goal of this work.

The analyses show that MELCOR 2.2 can model iPWR design and simulate severe accident scenarios with different levels of core degradation. One of the technical insights from this preliminary study was that natural circulation plays a significant role in the late phase of a severe accident when the core is uncovered.