Accuracy of Batan-3DIFF and MCNP6 codes for thermal neutron flux distribution at the irradiation position of RSG-GAS reactor

This paper presents the accuracy of the neutron diffusion and Monte Carlo methods in determining the axial thermal neutron flux in the irradiation facility. The calculations are compared with the activation method. The Au-foils are inserted close to the targets, the Low Enrichment Uranium (LEU) electroplating and TeO2. A new cell model for the thin layer of the LEU target is proposed. The MCNP calculation results are in very good agreement with the experimental core parameters, such as excess reactivity and total control rod worth. However, the neutron diffusion method code, Batan-3DIFF, has a higher relative difference of 10.14%. The proposed multi-zone homogenised model can improve the criticality calculation by 124% compare to the homogenised cell. For the axial thermal neutron distribution, the MCNP gives very satisfying results within the standard deviation of experimental results. However, the Batan-3DIFF code has an average relative difference of 7%.

Keywords: RSG-GAS reactor; Au-foils; thermal neutron flux; Batan-3DIFF; MCNP6.1.