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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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The 2025 edition is a reinvigoration of the withdrawn ANS-54.8-1988 of the same title.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.
J. H. Shaffer, W. R. Grimes, G. M. Watson, D. R. Cuneo, J. E. Strain, M. J. Kelly
Nuclear Science and Engineering | Volume 18 | Number 2 | February 1964 | Pages 177-181
Technical Paper | doi.org/10.13182/NSE64-A18316
Articles are hosted by Taylor and Francis Online.
In the conceptual two-region molten-salt breeder reactor, fissionable U233 will be recovered from the blanket as the decay product of Pa233. Since equilibrium concentrations of Pa233 would result in appreciable parasitic neutron absorptions, the advantages of thermal breeding could be realized to a greater extent by removing both Pa233 and U233 from the blanket mixture. Methods for recovering these materials from molten-fluoride mixtures by precipitation as oxides are presented. Small-scale experiments clearly indicated that it is possible to remove protactinium from molten-fluoride solutions by a process that appears to be surface precipitation of protactinium on beryllium oxide or thorium oxide particles. Protactinium was removed from molten mixtures of LiF-BeF2-ThF4 (67-18-15 mole %) by the addition of 1 to 2% by weight of solid beryllium oxide or thorium oxide. The removal efficiency was high when the initial concentration of protactinium was either in the range 1 to 2 ppb or 50 to 75 ppm. Uranium was successfully removed from solution in molten fluorides by use of a similar procedure. Approximately 2000 ppm uranium was precipitated from molten LiF-BeF2-ThF4 (67-18-15 mole %) by the addition of 3% by weight of beryllium oxide. Comparable results were also obtained using thorium oxide as the precipitant.
