SUMMARY

Current nuclear fuel cladding and structural materials struggle to survive the combined effects of temperature, stresses, and corrosion found in advanced reactor cores; and the development of new materials is expensive and time consuming. An attractive, low-cost, near-term solution is the multi-metallic layered composite (MMLC) material, a material design concept where the tasks of structural integrity and corrosion resistance are assigned to separate material layers. In this work, the operating envelopes of multiple advanced reactor design concepts will be computationally analyzed and potentially expanded through the implementation of MMLC fuel-cladding materials. First, reactor operating envelopes, defined as the set of neutronic, thermal-hydraulic, and thermo-mechanical reactor core conditions, will be identified with two computational tools: the SERPENT Monte Carlo neutron transport code for neutronic analyses, and the REX code for coupled thermal-hydraulic/thermo-mechanical analyses. REX, a provisionally patented reactor design tool developed specifically for this work, performs a sequence of core subchannel calculations; in which, fuel pin geometries are iteratively modified to force prohibitive thermal and fuel-clad contact stresses in precursory cladding materials. It then quantifies the thermo-mechanical response of MMLC fuel-clad materials with initially-limiting geometries as a function of corrosion-resistant layer width. The calculation sequence concludes with a thermal-hydraulic operating envelope expansion sequence; in which, the MMLCs that expand the reactor thermo-mechanical performance are tested at elevated coolant temperatures. The results of this work are quantified thermal-hydraulic and thermo-mechanical performance gains in lead-cooled and fluoride salt-cooled designs with MMLC fuel-cladding materials. A base thermo-mechanical design standpoint is determined for a sodium-cooled reactor core whose functional MMLC has yet to be identified. Finally, thermal-hydraulic performance gains for the intermediate heat exchanger of a compact fusion device are determined with a simplified version of the REX framework.