Inter-comparison of the recriticality temperature calculated by the KARATE-440 code system with real plant data

The possible return to power during the strong cooling down following a steam line break in the pressurized water reactors is a potential hazard. Among others, this transient can be characterized by the recriticality temperature which strongly depends on several three dimensional effects existing in the reactor core. This parameter can be evaluated by a best estimate neutronic code but an appropriate conservatism must be assured. Due to the nature of this transient the recriticality temperature is practically non-measurable and its uncertainty is hard to predict. Even so, some Russian exceptional measurements and appropriate core data referred in the literature can be found. In this paper our common practice is outlined, the measurements are presented and a simulation made by the Hungarian VVER-440 specific code system called KARATE-440 is described on the basis of the available plant data.