Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor (original) (raw)
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Elevated temperature microcantilever testing of fresh U-10Mo fuel
Journal of Nuclear Materials, 2019
Proliferation concerns are the driving force to reduce the enrichment of research reactor fuel from highly enriched fuel to lower levels of enriched fuel. One promising fuel alloy that would enable a reduction in enrichment without greatly diminishing fuel performance is Uranium with 10 wt % Molybdenum (U-10 wt-% Mo). While there is a large amount of data available on the microstructure and thermal properties of this new fuel type currently there is insufficient data available on the mechanical properties. Small scale mechanical testing techniques can be used to evaluate the mechanical properties of the asfabricated U-10 wt% Mo components at their intended operating temperature which should improve the accuracy of the input parameters employed by modelers. In this study the mechanical properties of a U-10 wt% Mo fuel meat sample taken from an as-fabricated fuel plate for a new fuel design were evaluated using in-situ scanning electron microscopy microcantilever testing under varying temperatures between room temperature and 200 C.
U-Si and U-Si-Ai Dispersion Fuel Alloy Development for Research and Test Reactors
Nuclear Technology, 1983
The reduction of the ^^^U enrichment from above 90% to below 20% for such fuels would lessen the risk of diversion of the fuel for nonpeaceful uses. Fuel alloy powder prepared with low-enrichment uranium (<20% "^U) is dispersed in an aluminum matrix, and metallurgically roll bonded within a cladding of 6061 aluminum alloy. Miniplates with up to 55 vol% fuel alloy (up to 7.0 g total U/cm^) have been successfully fabricated. Fifty-five of these plates have been or are being irradiated in the Oak Ridge Research Reactor. Three fuel alloys have been used in the ANL miniplates: UsSi (U-\-4 wt% Si), UsSi2 (U + 7.5 wt% Si), and "U^SiAl" (U + 3.5 wt% Si + L5 wt% Al). All are candidates for permitting higher fuel loadings and thus lower enrichments of ^^^U than would be possible with either UAl^ or UsOg, the current fuels for plate-type elements. A target loading of up to 7.0 g U/cm^ in the fuel zone was selected. To date the fabrication and irradiation results with the silicide fuels have been encouraging, and as an adjunct to the development effort, ANL is engaged in the early stages of technology transfer with commercial fabricators of fuel elements for research reactors. Continuing effort also involves the development of a technology for full-sized plate fabrication and the irradiation of miniplates to a burnup of-90% "^U depletion.
Low-enriched uranium-molybdenum fuel plate development
To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution.
Journal of Nuclear Materials, 2009
Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127 I beam up to an ion fluence of 2 Â 10 17 cm À2. Microscopy and mainly Xray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl 3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.
Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel
Journal of Nuclear Materials, 2016
Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions-such as thermal cycling-are some of the technological problems for this new fuel. In this paper, the effect of outof-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 C at a rate of 1 C/min (~125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.
Journal of Nuclear Materials, 2014
Microstructural changes in a set of commercial grade UO 2 fuel samples have been investigated using synchrotron based micro-focused X-ray fluorescence (l-XRF) and X-ray diffraction (l-XRD) techniques. The results are associated with conventional UO 2 materials and relatively larger grain chromia-doped UO 2 fuels, irradiated in a commercial light water reactor plant (average burn-up: 40 MW d kg À1). The lattice parameters of UO 2 in fresh and irradiated specimens have been measured and compared with theoretical predictions. In the pristine state, the doped fuel has a somewhat smaller lattice parameter than the standard UO 2 as a result of chromia doping. Increase in micro-strain and lattice parameter in irradiated materials is highlighted. All irradiated samples behave in a similar manner with UO 2 lattice expansion occurring upon irradiation, where any Cr induced effect seems insignificant and accumulated lattice defects prevail. Elastic strain energy densities in the irradiated fuels are also evaluated based on the UO 2 crystal lattice strain and non-uniform strain. The l-XRD patterns further allow the evaluation of the crystalline domain size and sub-grain formation at different locations of the irradiated UO 2 pellets.
Journal of Nuclear Materials, 2019
Corrosion behavior is an important consideration in the fuel qualification process. No prior study comprehensively evaluated the corrosion behavior of uranium-10 wt percent (wt%) molybdenum alloy for use in U.S. High Performance Research Reactors with the current composition, thermomechanical processing methods, dimension, cladding material, interface, research reactor water chemistry, temperature, and flow conditions. Hence, this work was performed to obtain baseline information about the corrosion behavior of DU-10Mo (depleted uranium-10 wt% molybdenum) alloy as a function of fabrication parameters (homogenization and hot rolling) using deionized water in a non-flow condition at 23 C and 75 C. Electrochemical and immersion corrosion testing were performed to determine the corrosion behavior of DU-10Mo hot-rolled foils. Microhardness testing, optical microscopy, and scanning electron microscopy were performed to evaluate the effect of fabrication parameters on DU-10Mo.