Irradiation performance of U-Mo monolithic fuel
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Argonne National Lab. (ANL), Argonne, IL (United States)
High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.
- Research Organization:
- Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- Grant/Contract Number:
- AC07-05ID14517
- OSTI ID:
- 1179370
- Report Number(s):
- INL/JOU-14-31685; PII: S1738573315301613; TRN: KR1503184084382
- Journal Information:
- Nuclear Engineering and Technology, Vol. 46, Issue 2; ISSN 1738-5733
- Publisher:
- Korean Nuclear SocietyCopyright Statement
- Country of Publication:
- United States
- Language:
- English
Web of Science
Advanced Postirradiation Characterization of Nuclear Fuels Using Pulsed Neutrons
|
journal | November 2019 |
Microstructure characterization and phase field analysis of dendritic crystal growth of γ-U and BCC-Mo dendrite in U–33 at.% Mo fast reactor fuel
|
journal | November 2017 |
Discontinuous Precipitation in U-10 wt.%Mo Alloy: Reaction Kinetics, Effect of Prior γ-UMo Microstructure, the Role of Grain-Boundary Misorientation, and the Effect of Ternary Alloying Addition
|
journal | June 2019 |
Similar Records
Interaction Layer Characteristics in U-xMo Dispersion/Monolithic Fuels
Effect of Interaction Layer and U-Mo Size on Pore Growth in U-Mo/Al Fuel
Related Subjects
ALUMINIUM
ALUMINIUM ALLOYS
DISPERSION NUCLEAR FUELS
URANIUM BASE ALLOYS
BINARY ALLOY SYSTEMS
RESEARCH REACTORS
DENSITY
IRRADIATION
MOLYBDENUM ALLOYS
ZIRCONIUM
FUEL PLATES
SWELLING
BURNUP
PERFORMANCE TESTING
PHYSICAL RADIATION EFFECTS
STABILITY
MATRIX MATERIALS
POWER-COOLING-MISMATCH ACCIDENTS
Mo2Zr Phase as Interaction Product
Monolithic fuel plate
radiation stability
U-10Mo/Zr/Al 6061