Institute for Neutron Physics and Reactor Technology (INR)

Core thermal hydraulic Analysis Methods

Motivation

The traditional way of performing thermal hydraulic core analysis is done using one dimensional system codes with single and two-phase flow models, which cover all heat transfer modes of the boiling curve. Using these codes, the core is represented by an “averaged” coolant channel that groups a number of fuel assembly together. Since many years, the use of 3D coarse mesh thermal hydraulic models started due to the implementation of 3D-models for the core in some system thermal hydraulic codes e.g. TRACE, RELAP5-3D, CATHARE-3D. In parallel, the use of subchannel codes increased in order to describe the core at fuel assembly level taking into account cross flow. Finally, there is an increased interest worldwide to perform subchannel level simulations of full cores by providing the pin power predicted either by pin-power reconstruction (PPR) or deterministic transport or Monte Carlo solver. Latter option, pave the way for direct prediction of local safety parameters at pin /subchannel level. These trends drive the development, improvement and coupling of subchannel codes with neutronic codes that are able to predict the pin and other safety parameters.   

Research goal

At KIT, the focus of this area is to improve the thermal hydraulic behaviour of reactor cores by applied methods with detail spatial discretization (subchannels: SubChanFlow) and better physical models for the energy and momentum equations (Porous media: TwoPorFlow) than the ones of 1D or 3D coarse mesh system codes.

Tools