Improved neutronic analysis for Cirus-PWL MOX fuels using 2D-transport theory lattice code WIMSD/4 and global multigroup 3D-diffusion code FINFOR

This paper discusses an improved analysis for the assessment of the fission power of a test fuel assembly containing MOX and thoria clusters in the pressurised water loop of Cirus reactor. Special emphasis is placed on the usage of WIMSD/4 2D transport theory code for the generation of nuclear data coupled with the multigroup 3D diffusion theory computer code FINFOR based on finite Fourier transformation technique for solving the diffusion equation in hexagonal geometry. The lattice parameters of the multiplying assemblies including Pressurised Water Loop (PWL) have been generated with the help of WIMSD/4 transport lattice theory code using 69 neutron energy group cross sections WIMS library. The DTF-IV (one dimensional transport theory code) has also been used to generate lattice parameters of non-multiplying assemblies. The computed power outputs of the test fuel assemblies are compared with available measured values. (author). 5 refs., 1 fig., 1 tab