Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

Abstract

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. The resulting interaction layer of 3–6 μm was measured after a 2-week exposure at 820 °C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1000 °C) and for Nuclear Reactor Severe Accident phenomenology study where the UO2 fuel could reach temperatures over 2800 °C and melt.