Power excursion of the training and research reactor of Budapest University

Abstract

A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the subchannel thermal-hydraulic core analysis code MI based on the momentum integral model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/MI for the accident analysis of the training and research reactor at the Budapest University of Technology and Economics (BME Reactor). An efficient and flexible cross-section generation procedure based on the WIMS code was developed and included in POWEX-K/MI. The code uses an implicit difference approach for the diffusion equations and thermal hydraulics. The neutronic part of the code was tested by simulating a start-up experiment performed at the BME reactor. The obtained results agree well with the experimental results. The code system was then applied to analysing power excursion accidents initiated by ramp reactivity insertions of $0.82 and $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.