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Title: Irradiation performance of U-Mo monolithic fuel

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,more » 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.« less
Authors:
 [1] ;  [1] ;  [1] ;  [1] ;  [1] ;  [1] ;  [1] ;  [1] ;  [2] ;  [2]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
  2. Argonne National Lab. (ANL), Argonne, IL (United States)
Publication Date:
OSTI Identifier:
1179370
Report Number(s):
INL/JOU-14-31685
Journal ID: ISSN 1738-5733; PII: S1738573315301613; TRN: KR1503184084382
Grant/Contract Number:
AC07-05ID14517
Type:
Accepted Manuscript
Journal Name:
Nuclear Engineering and Technology
Additional Journal Information:
Journal Volume: 46; Journal Issue: 2; Journal ID: ISSN 1738-5733
Publisher:
Korean Nuclear Society
Research Org:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 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