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Division Spotlight
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.
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.
J. F. Bates, M. K. Korenko
Nuclear Technology | Volume 48 | Number 3 | May 1980 | Pages 303-314
Technical Paper | Material | doi.org/10.13182/NT80-A32477
Articles are hosted by Taylor and Francis Online.
Irradiation-induced swelling in 20% cold-worked Type 316 stainless steel can be described by the use of a bilinear equation with three governing parameters. These parameters are R, a steady-state or linear swelling rate, τ, an incubation parameter denoting a fluence beyond which the linear, or high swelling, portion of the curve is attained, and a, a curvature parameter designating the degree of sharpness by which the equation curves from a region of low swelling to a region of higher swelling. This equation is intended for inclusion in the Nuclear Systems Materials Handbook and was developed with data extending to fluences around 16 × 1022 n/cm2(E> 0.1 MeV). The data set utilized includes first core Fast Flux Test Facility (FFTF) cladding specimens and specimens from several non-FFTF lots of cladding, in addition to supplemental data from an air-melted heat of steel. Heat-to-heat variations in swelling are significant in this material, and separate incubation parameters were developed for different lots of cladding.