Adaptation of multi-group version of DYN3D code to analytical studies on boiling water reactor with reduced neutron moderation - Analysis of reactivity accidents

Adaptation of multi-group version of DYN3D code was made to make analytical studies on boiling water reactor with tight fuel element lattice and fast neutron spectrum. For this purpose, the code was linked to nuclear cross section data generated by the lattice code IPPE-WIMS. Library of 5-group nuclear data was prepared for analytical studies on boiling water fast reactor. In order to confirm correctness of the preparation of the nuclear data and code performance for this type of reactor, series of test and demonstration calculations of steady state modes were made including calculations with base constants (ignoring thermal hydraulics feedbacks), as well as combined analysis of reactor neutronics and thermal hydraulics for the beginning and the end of the core lifetime. Analytical studies were performed on two reactivity accidents caused by ejection of highest-worth control rod and unauthorized withdrawal of control rods group in the reactor operating on rated power.