A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

This compilation of thermophyical and mechanical properties of certain metallic fuels is meant to be used as a common source of data in work related to the Integral Fast Reactor. Because research on these properties is an ongoing effort, this handbook must be continuously updated in order to provide the best data set to all involved in the IFR program. The use of cornmon source of properties will facilitate comparison of various analyses of fuel behavior performed within the program. It also rvill facilitate uncovering gaps and weaknesses in the data base, and thus enable better direction for future work on experimental properties work.

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An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin Finite Element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values.The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U-Pu-Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all available data covering 0 to 100% zirconium concentration and a zirconium concentration dependent linear heat rate solution derived from Monte Carlo N-Particle (MCNP) simulations were developed. An iterative calibration process was applied to obtain optimized diffusion coefficients for U-Pu-Zr metallic fuels. Optimized diffusion coefficients suggest relative improvements in comparison to previous reported values. The most influential or uncertain phase is found to be the gamma phase, followed by alpha phase, and thirdly the beta phase; indicating separate effect testing should concentrate on these phases.

Administration Under DOE Idaho Operations OfficeContract DE-AC07-05ID14517iii iv SUMMARY Based on data available in 2009, a decision was made to focus U-Mo monolithic fuel development and qualification efforts on a single fuel design. This fuel design consists of a U-10Mo (wt.%) monolithic fuel foil, a zirconium barrier layer applied to the faces of the foil, and 6061 aluminum cladding. This document updates the basis for the selection of that fuel system, drawing on additional fuel-performance data available as of May 2013. This update also applies recently developed GTRI (Global Threat Reduction Initiative) requirements for fuel qualification to the selection process.Because of its inherent gamma-phase stability and qualitatively better rolling behavior, U-10Mo was selected as the fuel alloy of choice, exhibiting reasonable uranium density and good irradiation performance. A zirconium barrier layer between the fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, greatly reducing fuel-cladding chemical interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot-and cold co-rolling with a Zr foil. The foils were subsequently bonded to the Al-6061 cladding by hot isostatic pressing or friction stir welding.Since the original down-select decision was made, additional information has become available on fuel performance and on the cost of the fuel system. Preliminary cost estimates indicate that U-Mo-Zr waste generated during fabrication is difficult to recycle and adds substantially to the production cost.