A new module for the AZtlan Nodal HEXagonal (AZNHEX) code, which is part of the AZTLAN Platform, was recently developed based on the Simplified Spherical Harmonics (SPL) scheme to deal with the challenges presented in small fast reactor cores, such as the China Experimental Fast Reactor (CEFR), with high leakage and significant scattering effects. For the verification and validation process, we generated nodal homogenized macroscopic cross-sections (XS) through a full heterogeneous core model using the stochastic code SERPENT and subsequently, these XS were employed in AZNHEX. To verify the SPL implementation, several mesh sensitivity exercises were performed demonstrating that the SPL module was implemented successfully. Furthermore, to validate the code with this new implementation, we modeled some exercises contained in the CEFR benchmark with AZNHEX and compared the results with the experimental data available. The final results show a great improvement compared with the original diffusion solver reducing the deviations significantly from experimental data. In conclusion, it is shown and discussed the relevance of improved numerical models (transport approximations instead of diffusion) for the deterministic calculations of small fast reactors.
A model of the Triga Mark III reactor of the National Institute for Nuclear Research (ININ) of Mexico was developed with the Monte Carlo codes Serpent and MCNP. The models were verified and validated by means of the experiments carried out as part of the starting tests to change the mixed fuel to Low Enrichment Fuels (LEU) of the TRIGA reactor core. The reactor data used in the V&V process consisted of fuel loading measurements, simulating the different stages of loading of fuel elements to the core to reach the reactor core criticality and the additional loading to achieve the reactivity excess to operate the reactor, as well as the evaluation of the shutdown margin reactivity and the control rods worth. The validated models constitute a trustworthy computational tool to analyse the most important neutronic core parameters as well as to have the numerical capabilities for fuel utilisation studies and analysis for extension of experimental facilities.
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