Analysis of the IAEA research reactor benchmark problem by the RETRAC-PC code

A basic approach to perform safety analysis of a nuclear research reactor consists in using deterministic methods to verify that the established acceptance criteria related to fuel integrity are fulfilled during all the stages of the facility lifetime. These methods should be validated against a large set of experimental and postulated transients. Since measured data are not easily available in the literature, the IAEA defined typical transients in a generic 10-MW MTR nuclear reactor core as a benchmark test for computational tools verification. In this framework, an assessment study of the coupled kinetic-thermal-hydraulic RETRAC-PC code is presented herein. The considered cases include the analysis of core dynamic under ramp positive reactivity insertion, and loss of flow transients. In general, the obtained results are satisfactory and agree with results obtained by other similar codes.