Analysis of the reactivity coefficients and fuel burnup for MNSR research reactor using DRAGON5/DONJON5 codes. (July 2019) - Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; El Mtili, Afaf
Abstract: The aim of this report is to analyze and validate a calculation, using new versions of the lattice DRAGON5 code (2D), the DONJON5 code (3D), and four cross-sectional libraries are ENDF/B-VII.0, ENDF/B-VII.1, JENDL4.0, and IAEA library. In this paper, the reactivity coefficients of a Miniature Neutron Source Reactor (MNSR) during burn-up changes are calculated. Two-dimensional DRAGON5 assembly transport and depletion calculations are used to compute the group constants of fuel assemblies. After a number of cell calculations in DRAGON5, the resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model. The full-core diffusion calculations are performed with three-dimensional DONJON5 code, from which both effective multiplication factor, excess reactivity, fuel temperature coefficient (FTC), moderator temperature coefficient (MTC), moderator density coefficient (MDC), and the channel power peaking factor are estimated, for different nuclear data libraries. Moreover, the dependency of coefficients with neutronic, thermal-hydraulic, and geometrical aspects is considered. The results of the calculation show a good agreement of the core reactivity coefficients with the previous publications. Highlights: Codes used include DRAGON and DONJON version 5. The MNSR core was analyzed to see the cross sections effects on reactivity coefficients. The FTC, MTC, and MDC were calculated using the DRAGON5/DONJON5 codes. Calculation of power peaking factor and fuel burnup time using the DRAGON5/DONJON5 codes. Good agreement between different nuclear data libraries when used by DRAGON5/DONJON5 codes.