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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.
Terrence A. Renner, Donald J. Raue
Nuclear Technology | Volume 42 | Number 3 | March 1979 | Pages 312-319
Technical Paper | Material | doi.org/10.13182/NT79-A32185
Articles are hosted by Taylor and Francis Online.
The permeability of tritium through Fe-2¼Cr-1 Mo steel (Croloy) steam generator material has been measured for the surface-clean metal and for the metal oxidized in steam at 755 K (482°C). The temperature dependence of tritium permeability has been measured for the unoxidized metal between 673 and 773 K; the apparent activation energy for permeation in this temperature range was found to be 39.14 ± 0.32 kJ/mol (9350 ± 80 cal/mol). After steam oxidation at 755 K and 0.2 MPa (2 atm) for about one month, the tritium permeation rate decreased by a factor of ≈150 relative to the clean metal. This reduction factor agrees very well with that measured for a sample of tubing used in Atomics International’s Modular Steam Generator experiments under typical liquid-metal fast breeder reactor steam generator conditions. These results may be used to calculate rates of tritium transport from the intermediate heat transport system sodium through the steam generator tubes and into the steam/water system. From this information, eventual release rates to the environment can be determined.