• A NUMERICAL TECHNIQUE FOR SOLVING COUPLED THERMALHYDRAULIC AND MULTI- ENERGY GROUP NEUTRON DIFFUSION EQUATIONS

Abstract

A sub-channel thermalhydraulic core analysis code CI based on the Channel Integral model is currently coupled to the three-dimensional neutronic code POWEX-K based on the neutron diffusion theory. This forms the integrated neutronic/thermalhydraulic code system POWEX-K/CI. The integrated model assessed against positive reactivity insertion transients in training and research reactors taking into account feedback effects due to coolant and fuel temperatures. An efficient and flexible cross-section generation procedure based on WIMS code was developed and included in POWEX-K/CI. The code system was then applied to analyzing power excursion accidents initiated by ramp reactivity insertions of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt occurs even if the safety rods do not fall into the core due to electrical, mechanical or any other reason.