Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Application of IRT-T research reactor as the potential solution of graphite radwaste disposal problem

3/23/2018 2018 - #01 Fuel cycle and nuclear waste management

Pavliuk A.O. Kotlyarevsky S.G. Bespala E.V. Bespala Yu.R.

DOI: https://doi.org/10.26583/npe.2018.1.09

UDC: 621.039.7

Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite elements of long-lived radionuclides, especially 14C, which may get involved in biological cycles because carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger of graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel following 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and on nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite placed in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of UGR designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the UGR. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor which was found to be equal to ~ 10 days was calculated. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including foreign-made reactors, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.


  1. Pavliuk A.O., Kotlyarevskij S.G., Bespala E.V. Experience of decommissioning of uranium-graphite reactor EI-2 in JSC «PDC UGR» Proc. «Radioactivity and radioactive elements in environment». Tomsk, 2016, pp. 508-512 (in Russian).
  2. Romenkov A.A. Results obtained: Handling with radioactive graphite during decommissioning of uranium-graphite reactors using technology of graphite oxidation in molten salt. Rosenergoatom. 2011, no. 3, pp. 32-35 (in Russian).
  3. Dunzik-Gougar M.L., Smith T.E. Removal of carbon-14 from irradiated graphite. Journal of Nuclear Materials. 2014, v. 451, pp. 328-335.
  4. LaBrier D., Dunzik-Gougar M.L. Identification and location of 14C-bearing species in thermally treated neutron irradiated graphites NBG-18 and NBG-25: Pre- and Post-thermal treatment. Journal of Nuclear Materials. 2015, v. 460, pp. 174-183.
  5. Liu J., Wang C., Dong L., Liang T. Study on the Recycling of Nuclear Graphite after Micro-Oxidation. Nuclear Engineering and Technology. 2016, v. 48, pp. 182-188.
  6. EPRI. Graphite Decommissioning: Options for Graphite Treatment, Recycling, or Disposal, including a discussion of Safety-Related Issues. Technical Report 1013091. 2006. 156 p. Available at: https://pdfs.semanticscholar.org/1367/38dccadbc420b7a112af9dd4c3b6885c6e5d.pdf (accessed Sept. 4 2017).
  7. IAEA. Disposal aspects of low and intermediate level decommissioning waste. Technical Report IAEA-TECDOC-1572. Vienna, IAEA Publ. 2007. 159 p. Available at: http://www-pub.iaea.org/MTCD/publications/PDF/TE_1572_companion_CD_web.pdf (accessed Sept. 4 2017).
  8. Vulpius D., Baginski K., Fischer C., Thomauske B. Location and chemical bond of radionuclides in neutron-irradiated nuclear graphite. Journal of Nuclear Materials. 2013, v. 438, pp. 163-177.
  9. Rublevskij V.P., Yatsemko V.N., Chanyshev E.G. The role of carbon14 in technogeneous irradiation of people. Moscow. IzdAT Publ., 2004. 197 p. (in Russian).
  10. Pageot J., Rouzaud J.-N., Gosmain L., Deldicque D., Comte J., Ammar M.R. Nanostructural characterizations of graphite waste from French gas-cooled nuclear reactors and links with 14C inventory. Carbon. 2016, v. 105, pp. 77-89.
  11. Smith T.E., McCrory S., Dunzik-Gougar M.L. Limited oxidation of irradiated graphite waste to remove surface carbon14. Idaho, Idaho State University Publ. 2012. 1776 p. Available at: http://dx.doi.org/10.5516/NET.06.2012.025 (accessed Sept. 4 2017).
  12. Dunzik-Gougar M.L., Smith T.E. Removal of carbon-14 from irradiated graphite. Journal of Nuclear Materials. 2014, v. 451, pp. 328-335.
  13. Fachinger J., von Lensa W., Podruhzina T. Decontamination of nuclear graphite. Nuclear Engineering and Design. 2008, v. 238, pp. 3086-3091.
  14. Dirk V., Kathrin B., Benjamin K., Bruno T. Thermal treatment of neutron-irradiated nuclear graphite. Nuclear Engineering and Design. 2013, v. 265, pp. 294- 309.
  15. Von Lensa W., Vulpius D., Steinmetz H.J., Girke N., Bosbach D., Thomauske B., Banford A.W., Bradbury D., Grambow B., Grave M.J., Jones A.N., Petit L., Pina G. Treatment and disposal of irradiated graphite and other carbonaceous waste. AtwInternational Journal for Nuclear Power. 2011, v. 57, pp. 263-269.
  16. Sibermann G., Moncoffre N., Toulhoat N., Bererd N., Perrat-Mabilon A., Laurent G., Raimbault L., Sainsot P., Rouzaud J.-N., Deldicque D. Temperature effects on the behavior of carbon-14 in nuclear graphite. Nuclear Instruments and Methods in Physics Research, Section B: Beam Interactions with Materials and Atoms. 2014, v. 332, pp. 106-110.
  17. Sach R.S., Williams W.J. The diffusion of 14C in nuclear graphite. Carbon. 1974, v. 12, pp. 425-432.
  18. Kane J.J., Karthik C., Ubic R., Windes W.E., Butt D.P. An oxygen transfer model for high purity graphite oxidation. Carbon. 2013, v. 2013, pp. 49-64.
  19. Poluektov P.P., Kashcheev V.A., Ustinov O.A., Musatov N.D., Yakunin S.A., Karlina O.K., Diordii M.N. Physicochemical aspects of reactor graphite incineration. Atomic Energy. 2014, v. 116, no. 2, pp. 105-109.
  20. Wickham A., Steinmetz H.-J., O’Sullivan P., Ojovan M.I. Updating irradiated graphite disposal: Project ‘GRAPA’ and the international decommissioning network. Journal of Environmental Radioactivity. 2017, v. 171, pp. 34-40.

irradiated reactor-grade graphite uranium-graphite reactor research reactor IRT-T