Electrochemical treatment of irradiated nuclear graphite

E. V. Bespala, M. V. Antonenko, D. O. Chubreev, A. V. Leonov, I. Yu Novoselov, A. P. Pavlenko, V. N. Kotov

Research output: Contribution to journalArticlepeer-review

3 Citations (Scopus)


The paper is dedicated to the issues of accumulation of irradiated reactor graphite and possible options for handling it. There are described the advantages of the electrochemical method of decontamination of radioactive waste widely used in the nuclear industry. The use of this method to reduce the potential hazard of irradiated reactor graphite is proposed. The processes occurring during the removal of radioactive contamination from the surface of graphite radioactive waste are described. An experimental setup is presented to assess the possibility of using the method of electrochemical decontamination of irradiated graphite. The results of the determining the electrolyzer current-voltage curve and the dissolution rate of the electrodes made of irradiated graphite are presented. According to the results of experimental studies, data on the decontamination coefficients of irradiated graphite were obtained for various radionuclides (60Co, 137Cs, 154Eu, 152Eu) in HNO3, H2SO4, H2O, H2O2, HNO3 + KMnO4, H2SO4 + KMnO4 under various process modes. The evaluation of temperature fields inside the electrolyzer was carried out. The dependence of the removal efficiency of long-lived 14C radionuclide on the total mass loss of irradiated graphite was obtained.

Original languageEnglish
Article number151759
JournalJournal of Nuclear Materials
Publication statusPublished - 1 Dec 2019


  • Decontamination
  • Electrochemical treatment
  • Electrolyzer
  • Irradiated graphite
  • Uranium-graphite reactor

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering

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