The United States faces a growing demand for High-Assay Low Enriched Uranium (HALEU) to support nuclear innovation and energy security. This work explores the feasibility of reprocessing and down-blending used naval nuclear fuel to create a supply of HALEU to decrease government liability while increasing nuclear innovation within the United States. Several nations successfully reprocess commercial spent nuclear fuel, and historically, the Department of Energy has also done so. The potential revenue of this project, at projected market costs, could supply operational costs for the government. Drawbacks lie in the capital costs of such a facility and the likely need to implement unproven technology. Ultimately, the reprocessing of Naval nuclear fuel presents a unique opportunity for the United States to reduce significant liabilities and invest in the country's energy future.
Plate-type fuel assemblies in high-flux research reactors are susceptible to flow blockage that may lead to fuel melting. Conventional system codes do not provide the local resolution needed for this problem, whereas full computational fluid dynamics (CFD) simulations are computationally expensive for routine accident analysis. This paper presents a computational framework for narrow rectangular plate-type fuel channels that covers the progression of a blockage event. A blockage scenario is mapped to a coupled three-dimensional solid and two-dimensional fluid model. Narrow-channel closures describe two-phase flow and critical heat flux (CHF), a dynamic porosity field updates flow accessibility, and a fully implicit scheme advances the coupled liquid, vapor, solid, and melt fields. These physical and numerical models are integrated into a broader workflow comprising model formulation, multiphysics coupling, numerical stabilization, separate-effect assessment, reactor-scale validation, and accident-scenario application. Comparisons with narrow-channel flow-boiling pressure-drop data, CHF experiments, melt/blockage and fuel-melting calculations, RA-6 heat-transfer measurements, and Japan Research Reactor-3 (JRR-3) reactor-scale results provide complementary assessment of the principal hydraulic, boiling, thermal, and material-evolution components of the framework. A transient based on the Tsinghua High Flux Reactor (THFR) then demonstrates how the same formulation follows a single modeled channel from intact forced convection through boiling and in-channel lateral redistribution to local overheating and melt transport.
