CFD simulation of critical heat flux in a rod bundle

The critical heat flux (CHF) condition is characterized by a sharp reduction of the local heat transfer coefficient which results from the replacement of liquid by vapour adjacent to the heat transfer surface. If the surface heat flux is the independent variable, the condition manifests itself as a sharp increase in surface temperature as the critical heat flux value is reached. The critical heat flux forms an important boundary for the performance of the heat exchange equipment. Determination of the critical heat flux is one of the key issues in nuclear reactor safety. This paper presents numerical simulations of boiling flow in a rod bundle with Departure from Nucleate Boiling (DNB) condition at the end of the middle rod. Large Water Loop CHF tests were used as a data set for our simulations. The Large Water Loop (LWL) is non-active pressurised-water equipment with technological and thermal parameters corresponding to those of PWR. The CHF experimental facility (a part of the Large Water Loop) has been designed for research into CHF in water flow through a bundle of electrically heated vertical rods. The critical conditions were determined under constant pressure, inlet water temperature and mass flux and for quasi steady-state - by gradually increasing the heat input. The rods are modelled by hollow tubes with direct heating of the wall. NEPTUNE-CFD code was used for numerical simulations. The computational domain covered a 30 deg. quasi-symmetric section of the actual channel. Simplified grid spacers were included in the domain. Calculations were performed with two-fluid approach with models for drag, lift, added mass and turbulent dispersion forces as well as for interfacial heat and mass transfer. Turbulent dispersion coefficient was based on void fraction gradient and on drag and mass forces. K-epsilon model was used for the prediction of the liquid turbulence, the flow of vapour was assumed to be laminar. Generalized wall heat-flux-splitting model was used to calculate production of vapour at the heated wall. This generalized model is an extension of the Kurul and Podowski model - it accounts for superheating of vapour under CHF conditions. A simple criterion based on the void fraction at the wall was used for the CHF prediction. Bubble mean diameter distribution in flow was calculated from one-group interfacial area transport equation with Yao's models for coalescence and break-up of the bubbles. Numerical simulations were performed for the several selected LWL test cases so as to find out whether the NEPTUNECFD can predict occurrence of the critical heat flux. After that, wall heat fluxes in simulation were increased or decreased so as to find out the interval of wall heat fluxes at which CHF condition occurs. So as to demonstrate the effect of grid spacers, results of one case are compared with simulation without grid spacers in the computational domain. The results show that NEPTUNE has potential for predicting the boiling flow up to CHF in the geometry of reactor fuel assembly. Presented work was done within 7. FP EURATOM NURISP project. NEPTUNE-CFD code is implemented in the NURESIM platform. (authors)