Thermal-hydraulic two-phase modeling of reactivity-initiated transients with CATHARE2 – Application to SPERT-IV simulation

Abstract

Reactivity-Initiated Accidents (RIA) in nuclear reactor cores are very complex multiphysic transients. They consist in a very fast power excursion in the core, leading to wall temperature excursions. Many experiments have shown that the heat exchanges coefficients during fast transients differ from those in steady-states. The thermal-hydraulic phenomenology of such transients remains complex and some experimental studies allowed to better understand and quantify the heat exchanges between the wall and the fluid, for both single and two-phase flows.The current study lies in a continuous effort of developing and validating a coupled neutronic and thermal-hydraulic code for the analysis of protected and unprotected transient behavior of reactors. More specifically, this study deals with the extension of the thermal hydraulic model of the CATHARE2 code to fast transient configurations. The extended CATHARE2 model is validated against experimental results from SPERT-IV. This reactor was driven mainly by the coolant density reactivity feedback. Thanks to that, many flow regimes can be observed in the core during the transient. That allows a more accurate evaluation of the adequacy of available transient two-phase flow models and correlations. The results, that represent the state of modeling, allow concluding that introduced models in CATHARE2 are able to simulate RIA tests such as SPERT-IV involving fast transient boiling.