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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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A new ANSI/ANS standard for liquid metal fire protection published
ANSI/ANS-54.8-2025, Liquid Metal Fire Protection in LMR Plants, received approval from the American National Standards Institute on September 2 and is now available for purchase.
The 2025 edition is a reinvigoration of the withdrawn ANS-54.8-1988 of the same title. The Advanced Reactor Codes and Standards Collaborative (ARCSC) identified the need for a current version of the standard via an industry survey.
Typical liquid metal reactor designs use liquid sodium as the coolant for both the primary and intermediate heat-transport systems. In addition, liquid sodium and NaK (a mixture of sodium and potassium that is liquid at room temperature) are often used in auxiliary heat-removal systems. Since these liquid metals can react readily with oxygen, water, and other compounds, special precautions must be taken in the design, construction, testing, and maintenance of the sodium/NaK systems to ensure that the potential for leakage is very small.
Allen R. Boynton, Robert E. Uhrig
Nuclear Science and Engineering | Volume 18 | Number 2 | February 1964 | Pages 220-229
Technical Paper | doi.org/10.13182/NSE64-A18321
Articles are hosted by Taylor and Francis Online.
An experimental method of measuring parameters peculiar to a two-region nuclear reactor is developed requiring the measurement of the cross-power spectrum between the outputs of the two reactor regions when a random reactivity input is given to one of the regions. Using bandpass filters and an analog computer, the cross-power spectrum between the outputs of the two regions in the University of Florida Training Reactor has been measured. These data indicate that the propagation of a disturbance from one region of the reactor to the other region may adequately be described in terms of neutron-wave phenomena and that the method may be used to determine the multiplication factor of each region.
