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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
Meeting Spotlight
Utility Working Conference and Vendor Technology Expo (UWC 2024)
August 4–7, 2024
Marco Island, FL|JW Marriott Marco Island
Standards Program
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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Latest News
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.
P. J. Fehrenbach, P. A. Morel, R. D. Sage
Nuclear Technology | Volume 56 | Number 1 | January 1982 | Pages 112-119
Technical Paper | Material | doi.org/10.13182/NT82-A32886
Articles are hosted by Taylor and Francis Online.
Measurement of fuel element diameters while the fuel is operating at power, in-reactor, has provided evidence of in-reactor fuel densification and relocation. The design and operation of the in-reactor diameter measuring rig used for these measurements are described. Diameter measurements were obtained from two fresh Zircaloy-clad UO2 elements containing fuel of 10.64 and 10.82 Mg/m3 density, respectively, at linear power outputs up to 61 kW/m. Similar measurements were also obtained from a 10.64 Mg/m3 density element after low power irradiation at 26 kW/m to a burnup of 75 MW- h/kg uranium. Results indicate that higher starting fuel density and prior irradiation both reduce the amount and rate of in-reactor fuel densification observed. Diameter measurements following reactor shutdowns, particularly on the higher burnup element, also indicate that fuel relocation can overcome diameter decreases due to fuel densification and restore pellet-clad interaction.