Use of Monte Carlo code MORET 5 for Research Reactor Safety analysis
Y. Chegrani, V. Tiberi, L. Heulers
RRFM 2012, Bruxelles, Belgique, March 18-22, 2012
This paper presents the use of the IRSN developed continuous-energy Monte Carlo code MORET 5 in French research reactor safety assessment. Three research reactor configurations have been studied. Two of these cases – ORPHEE and Jules Horowitz reactors (JHR) - focused on the assessment of accidental reactivity insertions involved in the analysis of BORAX-type accident (BORAX is an accidental transient due to a reactivity rush leading to a core melting and a steam explosion). ORPHEE study consisted in re-evaluating the worth of reactivity insertions caused by the neutron beam channels ruptures. The purpose of JHR study was to evaluate the worth of the reactivity insertion due to a control rod ejection. Both these results were widely used to direct the conclusions of the safety assessment. Basic calculations with MORET 5 allow obtaining keff evaluations for each configuration. But in order to fully extend the capabilities of this code to answer reactor safety specific issues, some developments have been performed in the 5.A.1 version of the MORET code. Thus, this version allows computing kinetic parameters, essential to the understanding of the reactivity accident progress. In order to validate this new functionality in MORET 5.A.1, the last chosen research reactor case deals with the validation of kinetic parameters models against measurements on the Brazilian reactor IPEN. The results obtained on IPEN allowed highlighting the needed improvements to increase the accuracy of kinetic parameters computation, in particular to better assess the power transient in the core. These calculations confirmed that MORET 5 is a convenient and reliable tool for the neutronic safety analysis of research reactors.