Journal of Radioanalytical and Nuclear Chemistry

, Volume 300, Issue 3, pp 1053–1059 | Cite as

Effects of compression molding on meltability of uranium dendrites for ingot consolidation in a pyroprocess

  • Jun-Hyuk Jang
  • Hee-Seok Kang
  • Ho-Se Lee
  • Sung-Jai Lee
  • Ki-Min Park
  • Jeong-Guk Kim


Compression molding was carried out to enhance the meltability of uranium dendrites. Uranium dendrites were successfully compressed, increasing their bulk density more than 9 times from 1.1 to 10 g/cm3. The average bulk density was about 8.7 g/cm3 which was almost half of the material density. The compressed dendrites were favorably melted at a temperature of 1,400 °C in an induction furnace. 3 kg of the compressed dendrites were consolidated into an ingot form, showing 96.7 % yield. About 3 % of dross was formed during the melting test in the form of fine powder which was characterized as a uranium oxide. This compression molding method was compared to the supplemental charge method in which uranium dendrites were poured into a molten metal pool produced from a uranium ingot. The capacity of dendrite melting was higher in the compression molding method than in the supplemental charge method. We consider that the higher capacity can be attributed to enhanced thermal conductivity as the bulk density was increased by the compression. These results suggest the high feasibility of the compression molding method for uranium melting in a pyroprocess at an engineering scale.


Uranium Dendrite Pyroprocess Compression molding Ingot casting 



This work was supported by the Nuclear Research & Development Program of the National Research Foundation (NRF), in a grant funded by the Korean Government.


  1. 1.
    Yoo JH, Lee BJ, Lee HS, Kim EH (2007) Investigation of pyroprocessing concept and its applicability as an alternative technology for conventional fuel cycle. J Korean Radioactive Waste Soc 5(4):283–295Google Scholar
  2. 2.
    Yoo JH, Hong KP, Lee HS (2008) A conceptual design study for a spent fuel pyroprocessing facility of a demonstration scale. J Korean Radioactive Waste Soc 6(3):233–244Google Scholar
  3. 3.
    Kwon SW, Kim JY, Ahn DH, Lee HS, Ahn HG (2010) A study on the evaporation of cadmium for the recovery of actinides from a liquid cathode. J Radioanl Nucl Chem 284:143–149CrossRefGoogle Scholar
  4. 4.
    Kwon SW, Park KM, Ahn HG, Lee HS, Kim JG (2011) Separation of adhered salt from uranium deposits generated in electro-refiner. J Radioanal Nucl Chem 288:789–793CrossRefGoogle Scholar
  5. 5.
    Kwon SW, Park KM, Jung Y, Ahn HG, Kim JG (2013) Development of an integrated sieve-crucible assembly for sequential operation of liquid salt separation and vacuum distillation. J Radioanal Nucl Chem 298:119–124CrossRefGoogle Scholar
  6. 6.
    Park KM, Kwon SW, Park SB, Kim JG (2012) The evaporation characteristics of LiCl–KCl eutectic salt from uranium deposits using batch type vacuum distiller with temperature slop of each zones. J Radioanal Nucl Chem 293:857–862CrossRefGoogle Scholar
  7. 7.
    Park KM, Kwon SW, Kim JG, Cho CH (2013) The solid–liquid separation characteristics of pure LiCl–KCl eutectic salt using different types of crucibles. J Radioanal Nucl Chem 295:1187–1193CrossRefGoogle Scholar
  8. 8.
    Kang HS, Jang JH, Lee YS, Lee H, Kim JG (2012) Development of engineering-scale ingot casting equipment for dendritic uranium deposit. Procedia Chem 7:758–763CrossRefGoogle Scholar
  9. 9.
    Jang JH, Kang HS, Lee YS, Lee H, Kim JG (2013) Development of continuous ingot casting process for uranium dendrites in pyroprocess. J Radioanal Nucl Chem 295:1743–1751CrossRefGoogle Scholar
  10. 10.
    Lee YS, Cho CH, Lee YS, Kim JG, Lee HS (2010) Uranium ingot casting method with uranium deposit in a pyroprocessing. J Korean Radioactive Waste Soc 8:85–89Google Scholar
  11. 11.
    Lee YS, Lee HS (2011) Ingot-casting apparatus using uranium deposits. US patent: US 2011/0056647 A0056641Google Scholar
  12. 12.
    Brunsvold AR, Roach PD, Westphal BR (2000) Design and development of a cathode processor for electrometallurgical treatment of spent nuclear fuel. In: Proceedings, 8th international conference on nuclear engineering, ICONE-8702, April 2–6, 2000, Baltimore, MD, USAGoogle Scholar
  13. 13.
    Westphal BR, Price JC, Vaden D, Benedict RW (2007) Engineering-scale distillation of cadmium for actinide recovery. J Alloy Comp 444–445:561–564CrossRefGoogle Scholar
  14. 14.
    Cho CH, Lee YS, Kim ES, Kim JG, Lee HS (2011) The reactivity with uranium of coating layers by the thermal spraying method. J Radioanal Nucl Chem 287:485–490CrossRefGoogle Scholar
  15. 15.
    Lee SH, Cho CH, Lee YS, Lee H, Kim JG (2010) Chemical reactivity of oxide materials with uranium and uranium trichloride. Korean J Chem Eng 27:1786–1790CrossRefGoogle Scholar
  16. 16.
    Totemeier TC, Mariani RD (1997) Morphologies of uranium and uranium–zirconium electrodeposits. J Nucl Mater 250:131–146CrossRefGoogle Scholar
  17. 17.
    Westphal BR, Marsden KC, Price JC, Laug DV (2008) On the development of a distillation process for the electrometallurgical treatment of irradiated spent nuclear fuel. Nucl Eng Technol 40(3):163–174CrossRefGoogle Scholar
  18. 18.
    Sakamura Y, Omori T, Inoue T (2008) Application of electrochemical reduction to produce metal fuel material from actinide oxides. Nucl Technol 162:169–178Google Scholar

Copyright information

© Akadémiai Kiadó, Budapest, Hungary 2014

Authors and Affiliations

  • Jun-Hyuk Jang
    • 1
  • Hee-Seok Kang
    • 1
  • Ho-Se Lee
    • 1
  • Sung-Jai Lee
    • 1
  • Ki-Min Park
    • 1
  • Jeong-Guk Kim
    • 1
  1. 1.Nuclear Fuel Cycle Process Technology Development DivisionKAERIDaejeonRepublic of Korea

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