Skip to main content
SpringerLink
Log in

In-reactor study of the friction characteristics of reactor materials

  • Articles
  • Published:
Soviet Atomic Energy Aims and scope

Conclusions

An apparatus was developed for measuring the frictional characteristics and was tested during in-reactor experiments. Apparently, the significant increase in the coefficient of static friction observed during the in-reactor irradiation process must be considered when evaluating the stress (deformed) state of the fuel elements during power fluctuations (jumps), i. e., in case of power increase after relatively long-term operation at a low power.

This is a preview of subscription content, log in via an institution to check access.

Access this article

Log in via an institution

Price excludes VAT (USA)
Tax calculation will be finalised during checkout.

Instant access to the full article PDF.

Institutional subscriptions

Similar content being viewed by others

Literature Cited

  1. Yu. K. Bibilashvili and G. I. Sukhanov, “Conference of the MAGATÉ specialists on extreme depletion of the fuel of aqueous power reactors,” At. Energ.,54, No. 2, 146–147 (1983).

    Google Scholar 

  2. “Barier fuel counters PCI,” Nucl. Eng. Int.,27, No. 333, 43 (1982).

  3. V. M. Shchavelin, A. V. Kostochka, A. M. Bolobolichev, et al., “Determining the friction characteristics of reactor materials,” At. Energ.,56, No. 3, 134–136 (1984).

    Google Scholar 

  4. A. M. Bolobolichev, A. V. Kostochka, A. A. Kuznetsov, and V. M. Shchavelin, “TREK-1 and TREK-2 devices for studying the frictional interaction of reactor materials under laboratory conditions,” in: Physics and Mechanics of Deformation and Fracture [in Russian], No. 10, Atomizdat, Moscow (1981), pp. 21–29.

    Google Scholar 

  5. A. M. Bolobolichev, A. V. Kostochka, A. A. Kuznetsov, and V. M. Shchavelin, “Intrareactor devices for studying the frictional interaction of the fuel and structural materials,”ibid. in:, pp. 29–38.

    Google Scholar 

  6. Handbook of Friction, Wear, and Lubrication [in Russian], Vol. 2, Mashinostroenie, Moscow (1978).

  7. W. Dienst, “Irradiation induced creep of ceramic nuclear fuel,” J. Nucl. Mater.,65, 1–8 (1977).

    Google Scholar 

  8. V. Fidleris, “Summary of the experimental results on in-reactor creep and irradiation growth of zirconium alloys,” At. Energ. Rev.,13, No. 1, 51–81 (1975).

    Google Scholar 

  9. É. S. Karakozov, Bonding of Metals in Solid Phases [in Russian], Metallurgiya, Moscow (1976), p. 264.

    Google Scholar 

Download references

Authors
  1. V. M. Shchavelin
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. A. V. Kostochka
    View author publications

    You can also search for this author in PubMed Google Scholar

  3. A. A. Kuznetsov
    View author publications

    You can also search for this author in PubMed Google Scholar

  4. I. S. Golovnin
    View author publications

    You can also search for this author in PubMed Google Scholar

  5. Yu. K. Bibilashvili
    View author publications

    You can also search for this author in PubMed Google Scholar

Additional information

Translated from Atomnaya Énergiya, Vol. 61, No. 3, pp. 175–178, September, 1986.

Rights and permissions

Reprints and permissions

About this article

Cite this article

Shchavelin, V.M., Kostochka, A.V., Kuznetsov, A.A. et al. In-reactor study of the friction characteristics of reactor materials. At Energy 61, 686–690 (1986). https://doi.org/10.1007/BF01129870

Download citation

  • Received:

  • Issue Date:

  • DOI: https://doi.org/10.1007/BF01129870

Keywords

Search

Navigation