VALIDATION OF MCNP FOR RBMK CRITICALITY CALCULATIONS

The Los Alamos National Laboratory Monte Carlo MCNP code is applied to critical experiments performed at the RBMK critical facility of the R ussian Research Center ''Kurchatov Institute,'' Moscow. The validation investigations are completed by whole-core criticality calculations o f experiments at the Smolensk Unit 3 nuclear power plant as part of th e start-up procedure. The geometric model exploits the powerful capabi lities of MCNP by precise representation of the fuel and different typ es of nonfuel channels, which add up to a detailed model of the critic al facility and the RBMK core. Continuous-energy cross-section tables are taken from the ENDF/B-IV and ENDF/B-VI libraries. As the most impo rtant uncertainty inherent to the experimental setup, the concentratio n of impurity isotopes in the graphite moderator is identified. Within the resulting error limits, the k(eff) and the void effect are well r eproduced with both cross-section libraries.