This study presents a combined computational-experimental approach to the design of radiation shields for a metallic Dual-Purpose Casks (DPC) intended for the transport and interim storage of spent fuel. The shield performance was simulated using ORIGEN 2.1 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 times. The results indicate that the isotope Cm-244 after 5 years of cooling, accounts over 98.2% of the spontaneous fission source, 29.9% of the (α, n) reaction mechanism, and 97.6% of the 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, in compliance 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.
Kiani et al. (Thu,) studied this question.