Abstract:
The research and development of advanced reactors has been pushed globally in recent decade and promotes the development of reactor physics codes. The neutron spectrum is complicate for some new reactors, especially for the special-use ones. In order to meet the requirements of the advanced reactors, it is necessary to develop the reactor analysis code with full neutron spectrum adapt ability. The full neutron spectrum code for advanced reactor simulation named FSAR has recently been developed at Nuclear Power Institute of China in order to meet the requirements of advanced reactor with large neutron energy range. The deterministic two-step calculation strategy based on the homogenization theory is utilized in FSAR to perform the reactor core neutronics analysis. Firstly, the two-dimensional lattice calculation is performed. The two-dimensional method of characteristics solver in hexagonal geometry for neutron transport calculation and the subgroup method with ultrafine groups for self-shielding calculation are utilized. For the typical assemblies, ultrafine-group cross sections and neutron flux will be determined and the few-group homogenized micro cross sections will be collapsed based on the flux-volume weight method and the principle of conservation of reaction rate. The single assembly problems with reflective boundary condition are used to determine the homogenized cross sections of materials. For better consideration of the strong space coupling in different geometric sizes due to different mean free path of neutrons, the super homogenization method and the leakage model are used as the homogenization technique. For the structural materials and other non-fuel materials, the super cell or assembly models are used to determine the SPH factors and the cross sections. Secondly, based on the multi-group cross sections generated by the lattice calculation, the three-dimensional whole core calculation is carried out by the core simulator. The SN method with triangular grid is applied as the solver of the transport equations. The three-dimensional multi-group neutron transport equation within the triangular prism grid will be calculated with isotropic fission sources and anisotropic scattering sources. The micro-depletion scheme is applied to simulate the core burn up and the Chebyshev rational approximation method is used to solve the depletion equation. In FSAR, the depletion chain containing 21 heavy isotopes and 49 fission products is provided for the advanced reactor core system. Preliminary verification of the resonance method was carried out. Several pin cell problems with different spectrum characteristics were selected as the verification problems. The results for the eigenvalue keff and for the absorption cross sections of 238U and 239Pu were compared to the reference results which were provided by the Monte Carlo code OpenMC. The results indicated that FSAR has good performances in dealing with the resonance self-shielding effect of the full-range spectrum problems.