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

• Published in: Journal of nuclear materials Vol. 512; no. C; pp. 252 - 267
• Main Authors: Parga, C.J., van Rooyen, I.J., Luther, E.P.
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 15.12.2018; Elsevier BV; Elsevier
• Subjects: Accident conditions; Assembly; Chemical reactions; Chemical reactors; Cladding; Conceptual Low-Enriched Uranium Fuel Element; Design; Design engineering; Enriched fuel reactors; Experimental; Fuel-Clad Chemical Interaction Evaluation; Fuel-clad interaction; Gas reactors; High temperature; High temperature gases; LEU fuel; Nuclear accidents & safety; Nuclear engineering; NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Nuclear fuels; Nuclear reactions; Nuclear reactors; Organic chemistry; Phenomenology; Reactors; Thermodynamic modeling; Thermodynamic models; Thermodynamics; TREAT reactor; Uranium; Uranium dioxide; Zircaloys (trademark); Zirconium; Zirconium base alloys; Thermodynamic modeling; TREAT reactor; LEU fuel; Experimental; Fuel-clad interaction
• ISSN: 0022-3115; 1873-4820
• DOI: 10.1016/j.jnucmat.2018.10.028

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. •The interaction between the TREAT LEU fuel and Zry-4 or Zr-1Nb alloys is minimal.•Thermodynamic modeling and experimental results provide good insight on FCCI process.•Zr-1Nb alloy has a slight advantage over the Zry-4 alloy performance.•Zry-4 forms an intermetallic layer at FCCI interface which is undesirable.•Zr-1Nb does not show segregation of elements near the FCCI interface which is preferred.