Coupling of Origen2.1 and Top-Mc for the Simulation and Experimental Validation of Spent Nuclear Fuel Cask Shields

This study presents a combined computational-experimental approach to the design of radiation shields for dual-purpose casks in dry storage systems. The shield performance was simulated using Origen2.1 and Top-Mc software and validated against laboratory experiments and MCNP simulations. Initially, using the ORIGEN2.1 code, the neutron and gamma fluxes resulting from nuclear mechanisms were calculated at different cooling time. The results indicate that Cm-244 element after 5 years from discharge, accounts over 98.2% and 29.9% and 97.6% of the spontaneous fission, (α,n) reaction mechanism and total neutron source, respectively. To evaluate and select the optimal radiation shield, dry storage casks containing 12 spent fuel assemblies were simulated using the Top-Mc software. By incorporating the radioactivity data obtained from ORIGEN2.1 calculations, a suitable composite shield for the transportation and dry storage of spent fuel was developed. The simulations demonstrated that a 10 cm-thick composite shield containing 20% boron carbide reduces the surface dose of the cask to 1.98 mSv/h, complying with the radiation safety requirements of SSR-6 standards. For benchmarking purposes, the designed neutron shield was fabricated and experimentally tested in the Tehran Research Reactor (TRR) against a thermal neutron spectrum. The shield achieved 99.69% thermal neutron absorption at a thickness of 0.99 cm. Furthermore, SEM imaging and elemental mapping analysis of the fabricated nanocomposite confirmed the uniform dispersion of filler particles within the polymer matrix. Experimental results were also simulated using MCNP code, revealing a maximum deviation of less than 7% between simulated and actual test conditions.