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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.
R. Carlander, S. D. Harkness, F. L. Yaggee
Nuclear Technology | Volume 7 | Number 1 | July 1969 | Pages 67-75
Material | doi.org/10.13182/NT69-A28387
Articles are hosted by Taylor and Francis Online.
Tensile properties of irradiated Type-304 stainless steel have been measured. These results have been correlated with microstructural observations obtained by optical and transmission electron microscopy techniques. The material studied was irradiated in a fast-neutron environment to a peak exposure of 4.8 × 1022 n/cm2 at temperatures ranging from 371 to 463°C in the EBR-II reactor. True yield stresses were observed to increase, and true uniform strains to decrease with both increasing neutron exposure and decreasing irradiation temperature for test temperatures <750°C. At 750°C no increases in true yield stresses over control values were noted while sharp decreases in true uniform strains were observed. It is suggested that some annealing of the microstructure occurs at this elevated temperature, allowing helium to be accumulated at grain boundaries. Microstructural examination by transmission electron microscopy revealed homogeneous distributions of polyhedral voids and Frank dislocation loops. Neither deject was observed to form on grain boundaries. It is suggested that the dislocation loop formation is primarily responsible for the increased strength of the irradiated material. Immersion density measurements are included. These results indicate that the peak void formation did not occur at the maximum flux position, thus indicating the importance of temperature to the phenomenon.