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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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.
D. Steiner
Nuclear Technology | Volume 9 | Number 1 | July 1970 | Pages 83-92
Reactor | Symposium on Theoretical Models for Predicting In-Reactor Performance of Fuel and Cladding Material | doi.org/10.13182/NT70-A28730
Articles are hosted by Taylor and Francis Online.
The neutronic behavior of fusion reactor blankets is discussed, and transport-theory calculations are presented for two blanket designs. The areas investigated are (1) tritium breeding, (2) nuclear heating, and (3) neutron irradiation effects within the vacuum wall of the blanket, i.e., neutron-induced (a) atom displacements and (b) helium and hydrogen production. The two blanket designs considered consist of niobium as the vacuum wall and structural material, lithium or lithium in combination with lithium-beryllium fluoride (called “flibe”) as the coolant, and graphite as the neutron moderator and reflector. The results indicate that the tritium breeding potential of both designs is promising. The results also show that the tritium-breeding and nuclear heating characteristics of the lithium-flibe blanket are inferior to those of the lithium blanket. The calculated atom displacement rates and production rates of helium and hydrogen within the vacuum wall are essentially the same for both blanket designs.