Abstract
Validation of the indigenously developed Monte Carlo program, MCkeff, is carried out by comparing the results of International Criticality Safety Benchmark Evaluation Project (ICSBEP) problems with that of standard legacy code MCNP (general-purpose Monte-Carlo N-particle transport code). Code-to-code criticality validation is essential before the usage of MCkeff to model various kinds of fissile systems. The Evaluated Nuclear Data Files (ENDF/B VII.I in ACE Format) used in MCNP is expended in the present study, thereby avoiding the differences in the results due to cross-section data. Thus, the deviations in the computed results will reflect only the differences arising out of differences in algorithms employed in these codes. Plutonium metal benchmark problems of different geometrical configurations were selected for validation purposes. All problems were simulated using MCkeff with the same inputs as in MCNP validation, and the computed results of keff (neutron multiplication factor) of various systems were compared. The results from MCkeff agree with that of MCNP, and the maximum deviation between them is less than 0.1%. Hence, the code MCkeff can be used to analyze other metal systems problems. Further validation work on the different fissile systems with other physical forms is in progress.
Published Version
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