Development of Neutronics Analysis Code System for Accelerator Driven Subcritical Reactor

The neutron flux density distribution of the accelerator driven subcritical system (ADS) is strongly anisotropic because of the external neutron source and thus neutron transport calculation is necessary. The Monte Carlo transport coupled with burnup calculation method is not efficient enough to be applied to the design work of ADS, and the traditional two-step reactor analysis method with the neutron diffusion method is not suitable. Therefore, the two-step ADS analysis code with the deterministic transport method was developed. The first step was carried out by the lattice code with the hyper-fine group resonance calculation method to provide the few-group homogenized parameters for the core calculation. The second step was executed by the three-dimensional discrete ordinates nodal transport calculation code (DNTR) in triangular meshes. The micro-depletion method was used for the burnup calculation of the core. The code was validated by a multi-cell burnup problem, the BN600 benchmark and the uniformly fuelled ADS model. The numerical results indicate that the code FMADS is reliable and efficient.