Izvestiya vuzov. Yadernaya Energetika

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

Physical characteristics of the large size sodium cooled fast reactors with advanced nitride and metal fuel

9/01/2015 2015 - #02 Physics and technology of nuclear reactors

Matveev V.I. Malysheva I.V. Bur’evskij I.V.

DOI: https://doi.org/10.26583/npe.2015.2.14

UDC: 621.039.526

Mixed nitride uranium-plutonium fuel is the most attractive and perspective fuel for fast sodium cooled reactors and this fuel is considered as a basic fuel for future commercial fast power reactors. However, a substantial increase of breeding using this fuel instead of oxide fuel is insufficient for satisfaction of new requirements main from which is minimization of reactivity margin for burn-up that determined by breeding in the core. The results of calculational studies to choose the optimum layout of the core of large fast sodium reactor meeting modern requirements are presented in the paper.

A metallic fuel for fast reactors has been considered since the beginning of their designing due to a high density, heat conductivity, and the minimum of dilute nuclei that provides maximum possible breeding.

The features of metallic fuel and the results of calculational studies of the use of metallic fuel in large fast sodium reactors with comparison to nitride fuel are presented in the paper. The conclusion drawn is that in the same layout of the core a metallic fuel is behind nitride fuel from the point of safety providing for the reactor.

The basic calculations have been conducted in the diffusion approximation on the basis of the known program complex TRIGEX.

References

  1. Zaboudko L.M., Mamaev L.I., Trufanov A.A. Analysis of calculated and experimental data for possible determine the causes depressurization of fuel rods with nitride fuel reactor BR-10 // Collection of papers from the seventh Russian conference on reactor materials science, Dimitrovgrad, 8-12 September 2003, p. 102 (in Russian).
  2. International conference ICAPP 2011, - Nice, France, May 2-5, 2011, Paper11340.
  3. Fast Reactor Databasa: 2006 Update. IAEA – TECDOC -1531.
  4. Adamov E.O., Zaboudko L.M., Matveev V. I. Rachkov V.I., Troyanov V.M., Khomyakov Y.S., Leonov V. N. A comparative analysis of the advantages and disadvantages of using metallic and nitride mixed uranium-plutonium fuel in fast reactors. Energy, Russian Academy of Sciences Publ., no. 2. 2015 (March-April 2015), pp. 3-15 (in Russian).
  5. Shimkevich A., Proshkin A., Sedov A. Promising dense fuel for power reactors. The collection REA. No. 10, October 2011 (in Russian).
  6. Zaboudko L. M. Operating Experience and problems of calculation justification dense fuels. Proceedings of the school-conference in Zelenograd, 08-11 November 2010. «Actual problems of development and production of nuclear fuel», Moscow, 2010, pp. 138-159 (in Russian).
  7. Seregin A.S., Kislitsina T.S., Tsibulya A.M. Abstract complex programs TRIGEX.04: Preprint SSC RF-IPPE -2846. Obninsk, 2000 (in Russian)..
  8. Kochetkov A.L. CARE Program – calculation of isotope kinetics, radiation and environmental characteristics of nuclear fuel during irradiation and aging: Preprint SSC RF-IPPE - 2431. Obninsk, 1995 (in Russian).
  9. Eliseev V.A., Zaboudko L.M., Malysheva I.V., Matveev V.I. Nitride fuel for the core of BN-1200 type advanced sodium cooled fast reactor. Atomic energy, v. 114, iss. 5, May 2013, pp. 266-271.
  10. Poplavsky V.M., Tsiboulya A. M., Khomyakov Yu.S., Matveev V.I., Eliseev V.A., Tsikunov A.G., Vasiljev B.A., Belov S.B., Farakshin M.R. Core Design and Fuel Cycle of Advanced Fast Reactor with Sodium Coolant. Atomic energy, v. 108, no.4, April 2010, pp. 206-211.
  11. Matveev V. I., Khomyakov Y.S. Technical physics sodium fast reactors with sodium coolant. Textbook for high schools. Under the editorship of Prof. interviewer Russian Academy of Sciences V.I. Rachkov. Moscow. Izdatel’skij dom MEI Publ., 2012, pp. 221 – 239 (in Russian).
  12. Eliseev V.А., Malysheva I.V., Matveev V.I., Egorov А.V., Maslov P.А. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel. // Global 2013, Salt Lake City, Utah. September 29-October 3, 2013, pp. 766-775.
  13. Bauer A.A., Cybulskis B., Green J.L. Mixed-nitride performance in EBR-II. Proc. of Int.Meeting on Advanced LMFBR Fuels, Tucson, October 10-13, 1977, p. 299.

fast sodium reactor nitride uranium-plutonium fuel metallic fuel fuel volume fraction reactivity margin for burn-up