Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

Abstract

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.