Validation of neutronics libraries through benchmarks and critical configurations of The Dalat Nuclear Research Reactor using low enriched uranium fuel by monte carlo method

Abstract

From evaluated data sources like ENDF, JENDL and JEFF, neutronics data libraries forMCNP computer code have been produced, including neutron scattering cross section library S (β,α) in thermal energy range, by using NJOY computer code. The evaluation and validation of these neutronics data libraries have been carried out through calculation of some parameters such as effective multiplication factor and reaction cross sections of benchmark problems, VVR-M2 fuel type as well as the critical configurations of the Dalat Nuclear Research Reactor loaded with low enriched Uranium fuel. After implementing about analysis and evaluation of the calculated results with abovementioned libraries, the library provides results that consistent with experimental data can be used in core and fuel management calculation for the Dalat Nuclear Research Reactor (DNRR).