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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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BWXT will scout potential TRISO fuel production sites in Wyoming
BWX Technologies Inc. announced today that its Advanced Technologies subsidiary has signed a cooperation agreement with the state of Wyoming to evaluate locations and requirements for siting a potential new TRISO nuclear fuel fabrication facility in the state.
J. Bentley, F. W. Wiffen
Nuclear Technology | Volume 30 | Number 3 | September 1976 | Pages 376-384
Technical Paper | Uranium Resource / Material | doi.org/10.13182/NT76-A31651
Articles are hosted by Taylor and Francis Online.
Vanadium-base alloys, V—10% Cr, V—20% Ti, and VANSTAR-7, alloys with potential for fusion reactor application, have been irradiated in the Experimental Breeder Reactor II in the temperature range of 400 to 800°C, mainly to a fluence of 1.5 × 1022 n/cm2 (>0.1 MeV). Swelling determined both from immersion density measurements and void distribution data obtained by transmission electron microscopy showed that the V—20% Ti was completely resistant to void formation for these irradiation temperatures and for the highest fluence achieved, 6 × 1022 n/cm2. Voids formed in both the V—10% Cr and VANSTAR-7 alloys, but only the V—10% Cr, irradiated at 690 and 805° C, showed technologically significant swelling, near 1%. Swelling in this alloy at lower temperatures and in VANSTAR-7 at all temperatures was below 0.1%. Dislocation structures were complex in all three alloys. In the V—20% Ti, the scale of the dislocation network coarsened with increasing irradiation temperature. In the other two alloys, the scale of the damage, both dislocation and void components, was similar for irradiation at 496 and 580°C, but coarsened considerably to produce similar structures for irradiations at 690 and 805°C. In many cases, detail of the microstructure was obscured by strongly diffracting zones that are believed to be impurity related. Of the three alloys examined, V—20% Ti possesses the greatest swelling resistance for the irradiation temperatures and fluences achieved and thus is judged to have the greatest potential for use in fusion reactors.