Browsing by Author "Wachs, DM"

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    Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding
    (Elsevier, 2016-10-01) Pasqualini, EE; Robinson, AB; Porter, DL; Wachs, DM; Finlay, MR
    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak). © 2016 Elsevier B.V.
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    Microstructural development in irradiated U-7Mo/6061 Al alloy matrix dispersion fuel
    (Elsevier, 2009-09-01) Keiser, DD; Robinson, AB; Jue, JF; Medvedev, P; Wachs, DM; Finlay, MR
    A U-7Mo alloy/6061 Al alloy matrix mini-dispersion fuel plate was irradiated in the Advanced Test Reactor and then examined using optical metallography and scanning electron microscopy to characterize the developed microstructure. Results were compared to the microstructure of the as-fabricated dispersion fuel to identify changes that occurred during irradiation. The layer that formed on the surface of the fuel U-7Mo particles during fuel plate fabrication exhibits stable irradiation performance as a result of the 0.88 wt% Si present in the fuel meat matrix. During irradiation, the pre-formed interaction layer changed very little in thickness and composition. The overall irradiation performance of the fuel plate to moderate power and burnup was considered excellent. © 2009, Elsevier Ltd.
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    Post irradiation examination of monolithic mini-fuel plates from RERTR-6 and 7
    (European Nuclear Society, 2007-03-11) Finlay, MR; Wachs, DM; Robinson, AB; Hofman, GL
    Successful qualification of the monolithic fuel is required for the conversion of the high performance research reactors and significant effort is being devoted to its development. The RERTR-6 experiment was designed to irradiate the first monolithic fuel mini-plates at moderate power density to moderate burn-up. The follow-on experiment, RERTR-7, aimed to irradiate monolithic fuel mini-plates at very high power density to high burn-up. It contained monolithic mini-plates fabricated by friction stir welding (FSW) and transient liquid phase bonding (TLPB). The post-irradiation of RERTR-7 indicates that porosity formation is occurring at the interface between the foil and the cladding in FSW plates. No porosity was observed in TLPB plates. The observations of delamination of FSW plates correlate with the initial mechanical testing results. A number of developments are being pursued on the fabrication front to address some of the observations and should be implemented in RERTR-9.