Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

Abstract

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Analyzing Simplified BWR Inherent Safety System using IAEA Generic Boiling Water Reactor Simulator

“…These emergency simulation results are slightly different from other studies. According to Mindaugas study, if the top of active fuel is uncovered for 18 seconds, it can rupture due to a ballooning [28]. Javier also showed that this emergency system could not stop the evolution of core degradation during a LOCA design basis accident progressing to a severe accident [29].…”

Study of nuclear reactor external vessel passive cooling using computational fluid dynamics

“…In addition, experimental measurements in passive safety systems have often been limited by the high cost, and technical difficulties in achieving the required spatial resolutions needed for detailed understanding, associated with the large scale containments and extended flow loops normally involved (Chu et al, 1997;Rouge, 1997;Theofanus et al, 1997). Numerical modelling has also been mainly restricted to onedimensional, best-estimate thermal hydraulic system codes (Park et al, 2016;Hu et al, 2018;Valincius et al, 2018) that, in adopting a one-dimensional approach, are not equipped to deal with the intrinsic three-dimensional, multi-scale nature of most natural circulation flow patterns.…”