Calculation of neutronic and kinetic parameters in Isfahan Miniature Neutron Source Reactor using Monte Carlo method and comparison with the results SAR
Document Type : Research Paper
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Abstract
Kinetic and neutronic parameters have an important role in reactors dynamic behavior analysis. Some of these parameters in nuclear reactors are such as effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt neutron lifetime (Lp). In the current work, analysis and calculation of the kinetic and neutronic parameters are performed using MCNPX code, slope fit and perturbation methods in Isfahan Miniature Neutron Source Reactor (MNSR). According to results, relative difference between the results of MCNPX code and the reference values in calculating of the reactivity and effective delayed neutron fraction are about 0.5% and 2.1%, respectively. The relative difference between the results of the slope fit and perturbation methods with the reference values in calculating of the prompt neutron lifetime are about 5.0%, 9.5% and 8.5%, respectively. Therefore, the results of this research show that the MCNPX code is suitable for calculating of the reactor kinetic parameters such as effective delayed neutrons fraction, while the perturbation method is a simple and convenient method for calculating of the prompt neutron lifetime.
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