Extension of Effective Cross Section Calculation Method for Neutron Transport Calculations in Particle-dispersed Media

概要

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A method for calculating effective microscopic and macroscopic cross sections for media containing randomly dispersed particles, which was originally derived by Shmakov et al., is improved to extend its applicability to a wider range of criticality calculations. The newly modified method can be applied to media containing more than one particle type. This technique is incorporated into a continuous energy Monte Carlo code MCNP and can be applied to a wide variety of criticality problems. The cause of an inadequacy of the original method for larger particles is investigated, and the accuracy is found to be improved by using an adjustment parameter. This method originally underestimated fission neutrons’ tracks in particles, which degrades calculation results for larger particles. A simplified method to implement the effect of the fission neutron tracks into MCNP is developed. We demonstrate that the new technique is successfully applied to MOX fuel rods with plutonium spots, fuel solution containing absorber particles, etc. The newly improved method can treat media containing particles with unequal diameters. However, a sample calculation shows the accuracy of criticality calculations becomes worse with increasing variation in particle diameters. The new method can also successfully perform a criticality calculation for media containing different particle compositions with an equal diameter.
社団法人日本原子力学会の論文
2006-01-25