EVALUATION OF HELICAL-CRUCIFORM FUEL ROD ASSEMBLIES FOR HIGH-POWER-DENSITY LWRs

摘要

Because of the immense capital costs associated with new nuclear construction, interest remains high in developing strategies to uprate existing light water reactors (LWRs) for higher power density and in raising core power density for next-generation LWR designs. Toward these goals, the helical-cruciform (HC) fuel rod assembly has been proposed. The HC fuel rod assembly is a self-supporting nuclear fuel configuration consisting of four-petaled, axially twisted fuel rods closely packed in a square array. Advantages over traditional fuel geometry include a larger surface-to-volume ratio and improved radial mixing characteristics. The self-supporting nature of the assembly obviates the need for grid plates, improving core hydraulics. Past studies have identified these and other benefits of HC fuel rod geometry and have adapted its shape and design to LWR fuel assemblies for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. However, because of a lack of suitable thermal-hydraulic correlations to capture HC rod bundle flow behavior, this work fell short of a complete assessment of the potential. Recent progress has been made in this regard due to the empirical development of specialized hydraulic and lateral mixing correlations for HC rod geometry. As a result, advanced LWR core designs taking advantage of the HC fuel rod assembly can be reexamined with a greater degree of precision and confidence. For a BWR core using HC rod assemblies, applying the new HC rod bundle correlations to subchannel models uncovered a need to increase the hydraulic diameter of the tight side and corner subchannels, to prevent flow starvation. Small protrusions were added to the assembly box side at axial locations corresponding to each rod quarter-twist to act as spacers. This prompted a slight redesign of the rod cross-sectional shape. Likewise, the central water rod region was adjusted to maintain the reference hydrogen-to-uranium atom ratio. With these changes, subchannel models predicted a 24% allowable power uprate for the 200-cm twist pitch HC core, in comparison to a reference BWR with traditional fuel. The uprate is accomplished assuming a fixed-core power-to-flow ratio. In comparison, modeling showed that a PWR core employing HC fuel rod assemblies may allow power uprates up to 47%, for a fixed power-to-flow ratio. One major difference from the BWR case is that subcooled critical heat flux (CHF) levels rise with increasing coolant mass velocity, opposite the trend for saturated CHF conditions. However, subcooled CHF is also known to be more sensitive to locally peaked heat flux, which was not explicitly modeled in these simulations. Power density gains claimed here will be ultimately dependent on the degree to which the HC rod's twist disrupts nascent pockets of vapor as subcooled CHF limits are approached; this effect should be further investigated experimentally.