Reactor physics assessment of candidate accident-tolerant cladding concepts for long-life civil nuclear marine propulsion cores

In this reactor physics study, we examine the neutronic performance of accident-tolerant fuel (ATF) claddings - Austenitic type 310 stainless steel (310SS), ferritic Fe-20Cr-5Al (FeCrAl), advanced powder metallurgic ferritic (APMT), and silicon carbide (SiC)-based materials - as alternative cladding materials compared with Zircaloy-4 (Zr) cladding. The cores considered in this study use 15% enriched homogeneous UO2 fuel and a constant cladding coating thickness of 655 μm was assumed throughout this study. We use the WIMS reactor physics code to analyse the associated achievable discharge burnup, changes in reactivity coefficients, and spectral variations to compare the different neutronic cases for candidate cladding materials. The results show that the fuel with 310SS cladding exhibits a ∼13% discharge burnup penalty compared to Zr due to the presence of a very high nickel (Ni) concentration; Ni has a thermal neutron absorption cross-section about twice that of iron (Fe). In addition, the high neutron absorption cross-sections of Fe in the FeCrAl and APMT claddings also lead to a ∼10% discharge burnup penalty. It was observed that the fuel with SiC cladding can achieve a ∼1% higher discharge burnup compared to Zr cladding due to the low thermal neutron absorption cross-sections of its constituents along with the presence of a softer neutron spectrum arising from reduced plutonium breeding throughout the cycle. The highest achievable discharge burnup with the SiC cladding provides the highest uranium utilisation, making this cladding the natural choice to take forward in this study. Reactivity feedback parameters (moderator and fuel temperature coefficients) were calculated throughout the cycle and identical responses were found when fuel, coolant temperature and coolant properties were perturbed for each cladding material. By splitting the pellet into 6 equal annular areas, the relative fission power as a function of radius was evaluated. It was observed that the claddings with lower capture cross-section (SiC and Zr) exhibit higher relative fission power at the pellet periphery when compared to FeCrAl, APMT and 310SS. Finally, it can be concluded from this study that, when modeling SiC, a standard cladding thickness could be implemented with marginally less enriched uranium. On the other hand, in order to overcome the burnup penalty while using an ironbased alloy cladding (FeCrAl, 310SS and APMT), it is recommended to reduce the cladding thickness and slightly increase the fuel enrichment in order to match the cycle length achieved with Zr