Radiological performance of cemented FMA-VC radioactive waste packages for transport: A PHITS Monte Carlo study

The transport of low- and intermediate-level short-lived radioactive waste (FMA-VC) necessitates stringent control of external radiation exposure to ensure compliance with international radiation protection regulations. In this study, a Monte Carlo-based dosimetric evaluation of cemented FMA-VC waste packages under transport conditions was performed using the PHITS (Particle and Heavy Ion Transport System) code. Two representative configurations (F3-1-03 and F3-5-05) were modeled with realistic cylindrical geometries incorporating the cement matrix, steel drum containment, and optional high-density shielding layers. External gamma dose rates were calculated at the package surface and at distances of 1 m and 2 m, in accordance with IAEA SSR-6 requirements. The attenuation performance of additional shielding materials, namely lead and tungsten, was also quantified. The results indicate that the intrinsic shielding provided by the cement matrix and steel drum ensures substantial gamma attenuation, with surface dose rates in the range of 10^-4-10^-3 mSv·h^-1, well below regulatory limits. Supplementary shielding further reduces dose rates to near-background levels, with tungsten demonstrating superior attenuation efficiency. These results confirm the radiological safety and regulatory compliance of the studied packages under routine transport conditions. Furthermore, the proposed modeling framework underscores the reliability and predictive capability of Monte Carlo simulation as a robust tool for safety assessment and optimization in radioactive waste transport.