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Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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. M. Beeston, R. R. Hobbins, G. W. Gibson, W. C. Francis*
Nuclear Technology | Volume 49 | Number 1 | June 1980 | Pages 136-149
Technical Paper | Nuclear Power Reactor Safety / Fuel | doi.org/10.13182/NT80-A32515
Articles are hosted by Taylor and Francis Online.
Uranium aluminide powder production, fuel plate fabrication development, and irradiation performance of more than 1700 fuel elements during 10 yr of operational service at Idaho National Engineering Laboratory are discussed. The UAlx dispersion fuel system has performed well in extended service in the high flux test reactors. The anticipated benefits of the powder dispersion form—accommodation of fission products in deliberate voidage, structural tolerance of fission gas, and dispersion of burnable poisons—have been realized. The operating limit for the Advanced Test Reactor fuel elements is presently set at 2.3 × 1021 fiss/cm3 of core—a burnup of >500 000 MWd/MTU. The growth or swelling of uranium aluminide fuel plates at up to 2.4 × 1021 fiss/cm3 is proportional to the fission density, but the proportionality constant depends on the temperature, core porosity, and fuel loading with 93% enriched uranium. For a fuel loading of 4.3 × 1021 U atoms /cm3, the growth corresponds to 0.11% per % burnup. The blister test as a criterion for impending fuel plate failure due to swelling appears adequate, and the blister temperature at fission densities of 2.7 × 1021 fiss/cm3 of core is ∼720 K.