Calculation and analysis of self-shielded cross sections for the high-fidelity neutronics calculation of fast reactors

A new high-fidelity neutronics analysis tool for fast reactor calculations is built by coupling the lattice code TULIP, the SN code HYDRA and the high-fidelity neutron transport code NECP-X. TULIP is employed to generate multigroup effective cross sections for assemblies based on the 0-D or 1-D cylindrical model. HYDRA utilizes a reactor core-reflector model to perform group condensation for reflector. The cross sections are passed to NECP-X to conduct a whole-core simulation with explicit geometry description. Verifications were conducted based on the Superphénix 2-D, MET-1000 2-D and JOYO MK-I 3-D core problems. The reference solutions were obtained through Monte Carlo calculations with NECP-MCX. The difference of keff was at most 298 pcm for 2-D problems and did not exceed 55 pcm for 3-D problem. The root mean square error of assembly power was a maximum of only 1 %. These results prove the capability of this method in fast reactor calculations.