TREAT reactor LEU fuel-clad chemical interaction empirical modeling analysis

Conference · Fri Jul 01 00:00:00 EDT 2016

OSTI ID:22764103

Parga, C. J.; Van Rooyen, I. J. ^[1]; Glazoff, M. V. ^[2]; Luther, E. P. ^[3]

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1. Fuel Performance and Design, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)
2. Advanced Process and Decision Systems, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)
3. Materials Science and Technology, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

DOE is converting TREAT (Transient Reactor Test) facility from its existing highly enriched uranium (HEU) core to a low-enriched uranium (LEU) core. In order to assess the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operational limits, planned transients tests, and reactivity accident scenarios, a combination of experimental testing and thermodynamic modeling was performed to predict the expected chemical interactions among fuel and cladding chemical constituents. Thermodynamic calculations were then validated with empirical data from experiments that emulate TREAT's expected upper operational limits. Pellet samples composed of LEU oxide powder dispersed in a graphite matrix had intimate contact with zirconium-based alloy cladding. The samples were subjected to long-term isothermal heating under high vacuum. Specimen characterization consisted of scanning electron microscopy, x-ray diffraction analysis, and x-ray tomography. ThermoCalc software and Ellingham's diagrams were used for thermodynamic calculations. The X-ray tomography and SEM analysis of the pellets showed a homogeneous distribution of UO{sub 2} particles within the graphite matrix with isolated larger UO{sub 2} agglomerates. The SEM analysis showed the formation of a lamellar structure between the large UO{sub 2} agglomerate and the graphite matrix. This is an indication that diffusion is taking place at a rather moderate temperature and reaction time during the pellet fabrication. Secondary phases, with high Al and Mg content, were detected inside a UO{sub 2} agglomerate using SEM in SE/EBSD mode and chemical analysis with EDS. LEU feedstock impurities, as Al and Mg, seem to promote the chemical reactivity between UO{sub 2} particles and graphite matrix. XRD analysis of a pellet sample showed the presence of UO{sub 2} and graphite, but no other crystalline phases were detected with this technique. Thermodynamic analysis through the use of Ellingham diagrams showed that within TREAT's operational to reactivity accident temperature (278-820 Celsius degrees), UO{sub 2} can be reduced to U metal by Zr due to a more favorable ΔG for the oxidation of Zr.

• Research Organization: American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22764103
• Resource Relation: Conference: TOP FUEL 2016: LWR fuels with enhanced safety and performance, Boise, ID (United States), 11-15 Sep 2016; Other Information: Country of input: France; 20 refs.; This record replaces 50007249; Related Information: In: TOP FUEL 2016 Proceedings| 1670 p.
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
36 MATERIALS SCIENCE
42 ENGINEERING
BACKSCATTERING
CHEMICAL ANALYSIS
CLADDING
DIAGRAMS
ELECTRON DIFFRACTION
FUEL PELLETS
GRAPHITE
HIGHLY ENRICHED URANIUM
NUCLEAR FUELS
OXIDATION
POWDERS
SCANNING ELECTRON MICROSCOPY
SIMULATION
T CODES
THERMODYNAMICS
TOMOGRAPHY
TREAT REACTOR
URANIUM DIOXIDE
X-RAY DIFFRACTION
ZIRCONIUM BASE ALLOYS