Izvestiya vuzov. Yadernaya Energetika

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

Interaction of Substitutional and Interstitial Atoms with each other, and with Radiation Defects in Alloys V−Fе under Irradiation with Fast Neutrons

6/05/2024 2024 - #02 Modelling processes at nuclear facilities

Folomeev V.I. Ganina S.M. Astakhova N.E.

DOI: https://doi.org/10.26583/npe.2024.2.16

UDC: 621.039.534:541.15

There is a substantial amount of experimental data that confirm the peculiarity of the oxide fuel behavior during the first hours after a reactor reaches its power. With an increase in temperature during said period, oxide fuel pellets crack because of a significant temperature gradient. Further developments occur due to the accumulation and redistribution of fission products, and manifest as changes to the fuel matrix porosity and the formation (or diameter increase) of a central hole. The zones that differ in their microstructure, density and thermal conductivity are formed along a fuel pellet radius. As the oxide fuel composition and restructuring result in its thermal conductivity change it’s important to pay attention to the formulas used in the stress-strain state calculations of a fuel element.

The methodology for calculating changes in oxide fuel porosity and a pellet’s internal diameter is proposed and based on the published research papers dedicated to the study of oxide fuel properties and behaviour during the first hours after a reactor reaches its power. Тhe methodology was tested using real experimental data on porosity redistribution along a fuel pellet radius.

A presence and a size of the pellet’s inner hole, as well as changes to the fuel matrix porosity, have a noticeable effect on the maximum temperature value. Taking into account a pellet structure evolution while performing computational modelling of fuel rods operation under irradiation allows assessing the fuel element’s operability more accurately.

The proposed methodology can be used in computer codes designed to calculate the stress-strain state of cylindrical fuel elements of fast reactors.


  1. Freund D. Fast reactor fuel structure evolution under irradiation. Kerntechnik. 1990, volume 55, issue 6, pp. 350 – 357. DOI: https://doi.org/10.1515/kern-1990-550612
  2. Degaltsev Y.G., Ponomarev-Stepnoy N.N., Kuznetsov V.F. Behaviour of high-temperatute nuclear fuel under irradiation. Moscow, Energoizdat, 1987, 208 p. (in Russian).
  3. Kinev E.A. The structure of pellet oxide fuel and its corrosive effect on the fuel pin clad of the BN-600 reactor. Izvestiya vuzov. Yadernaya energetika. 2011, no 1, p. 169 – 176. Available at: https://static.nuclear-power-engineering.ru/journals/2011/01.pdf (accesed Mar.15, 2024) (in Russian).
  4. Kozlov A.V., Kinev E.A., Tsygvintsev V.A. Post-reactor studies of oxide fuel after operating in the BN-600 reactor. VANT. Series Materials science and new materials. 2004, 2 (63), pp. 173 – 180 (in Russian).
  5. Olander Donald R. Fundamental aspects of nuclear reactor fuel elements. Published by Technical Information Center, Office of Public Affairs Energy Research and Development Administration. The United States of America. 1976. DOI: https://doi.org/10.2172/7343826
  6. Parrish R.J., Cappia F., Aitkaliyeva A. Comparison of radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products. Journal of Nuclear Materials. 2020, vol. 531, pp.1 – 8. DOI: https:// doi.org/10.1016/j.jnucmat.2020.152003
  7. Cappia F., Tanaka K., Kato M., McClellan K., Harp J. Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA. Journal of Nuclear Materials. 2023, vol. 533, pp. 1 – 14. DOI: https://doi.org/10.1016/j.jnucmat.2020.152076
  8. Venkiteswaran C.N., Jayaraj V.V., Ojha B.K., Anandaraj V., Padalakshmi M., Vinodkumar S., Karthik V., Vijaykumar Ran, Vijayaraghavan A., Divakar R., Johny T., Joseph Jojo, Thirunavakkarasu S., Saravanan T., Philip John, Rao B.P.C., Kasiviswanathan K.V., Jayakumar T. Irradiation performance of PFBR MOX fuel after 112GWD/t burn-up. Journal of Nuclear Materials. 2014, vol. 449, pp. 31 – 38. DOI: https://dx.doi.org/10.1016/j.jnucmat.2014.01.045
  9. Magni A., Barani T., Del Nevo A., Pizzocri D., Staicu D., Van Uffelen P., Luzzi L. Modelling and assessment of thermal conductivity and melting behaviour of MOX fuel for fast reactor applications. Journal of Nuclear Materials. 2020, vol. 541, pp. 1 – 13. DOI: https://doi.org/10.1016/j.jnucmat.2020.152410
  10. Bonev P., Chauvin N., Staicu D., Dahms E., Montagnier G., Papaioannou D., Dumas J-C., Boukhris I., Viallard I., Lainet M., Lamontagne J., Hanifi K. New recommendation for the thermal conductivity of irradiated (U,Pu)O2 fuels under fast reactor conditions/ Comparison with recent experimental data. Journal of Nuclear Materials. 2023, vol. 577, pp. 1 – 11. DOI: https://doi.org/10.1016/j.jnucmat.2023.154326
  11. Lackey W.J., Homan F.J., Olsen A.R. Porosity and actinide redistribution during irradiation of (U, Pu)O2. Nuclear Technology. 1972, vol. 16, pp. 120 – 142. DOI: https://doi.org/10.13182/NT72-A31181
  12. Philipponneau Y. Thermal conductivity of (U,Pu)O2-x mixed oxide fuel. Journal Nuclear Materials. 1992, vol. 188, pp. 194 – 197. DOI: https://doi.org/10.1016/0022-3115(92)90470-6

oxide fuel porosity restructuring fuel pin fuel pellet experimental data stress-strain states (SSS) for fuel rods

Link for citing the article: Folomeev V.I., Ganina S.M., Astakhova N.E. Interaction of Substitutional and Interstitial Atoms with each other, and with Radiation Defects in Alloys V−Fе under Irradiation with Fast Neutrons. Izvestiya vuzov. Yadernaya Energetika. 2024, no. 2, pp. 202-212; DOI: https://doi.org/10.26583/npe.2024.2.16 (in Russian).