Neutronic performance of accident-tolerant fuel configurations in a PWR-type small modular reactor assembly: a Monte Carlo study using OpenMC

The Fukushima-Daiichi accident renewed interest in accident-tolerant fuel (ATF) systems that offer extended safety margins during loss-of-coolant events. This paper presents a Monte Carlo eigenvalue assessment of four fuel-cladding configurations in a BEAVRS 17 × 17 PWR-type assembly at beginning-of-life, using OpenMC 0.15.3 with the ENDF/B-VIII.0 nuclear data library. The assembly model uses a BEAVRS-based 17 × 17 PWR assembly model with light-water coolant; soluble-boron-free SMR designs require separate spectral and control analysis. The configurations are: (A) UO 2 /Zircaloy-4, (B) U 3 Si 2 /Zircaloy-4, (C) UO 2 /SiC, and (D) U 3 Si 2 /SiC. Forty eigenvalue calculations characterize k∞ across 3.5–5.0 wt% 235 U, the Doppler temperature coefficient (αD), the moderator temperature coefficient (αM), neutron spectrum, four-factor decomposition, and pin-power distribution at nominal conditions (4.5 wt%, T fuel = 900 K, T cool = 573 K). SiC cladding provides a reactivity gain of +332 to +344 pcm through reduced parasitic absorption and spectral softening, partially offsetting spectral hardening from U 3 Si 2 fuel. U 3 Si 2 incurs a reactivity penalty of −635 to −704 pcm from spectral hardening and carries a lower melting point (∼1,665 °C versus ∼2,865 °C for UO 2 ), reducing thermal safety margins during transients. All four configurations maintain strongly negative αD (−1.543 to −1.590 pcm K −1 ) and αM (−26.6 to −30.3 pcm K −1 ), confirming inherent neutronic safety feedback. The pin-power peaking factor for Config D is F q = 1.062. These results establish a quantitative neutronic baseline for ATF deployment in PWR-type SMR assemblies; a complete safety evaluation requires thermal-hydraulic and depletion assessments.