Preliminary neutronic analysis of alternative cladding materials for APR-1400 fuel assembly

Abstract This paper presents a preliminary neutronics analysis for Accident Tolerant Fuel (ATF) cladding materials for standard APR-1400 reactor A0 fuel assembly. AMPT, FeCrAl, 304 and 310 stainless steels, and SiC were considered and compared with the original Zircaloy-4 cladding material. A parametric evaluation was done for fuel and cladding materials to confirm the geometry requirements to achieve the end-of-cycle fuel reactivity, and the results were compared with the standard APR-1400 reference fuel-cladding system. Serpent Monte Carlo reactor physics code version 2.31 was utilized to evaluate the associated neutronics penalty when different cladding materials were used such as, fuel reactivity, thermal neutron spectrum, plutonium and isotopic inventory evolution, and fuel assembly linear pin power distribution. In the examined cases, fuel enrichments were fixed, and cladding thickness and pellet diameter were changed for zircaloy, cases 1 – 4. As a result, thermal neutron absorption cross section varies between materials due to higher absorbing cross section leads to hardening of the neutron spectrum in the fuel-cladding system. That leads to a slight increase in the production of actinides in the inventory such as plutonium.