Atomic Energy

, Volume 92, Issue 5, pp 367–372 | Cite as

Some Problems of Large-Scale Safety Test Stands

  • V. N. Smolin
  • V. P. Shishov
  • V. I. Grachev
  • A. I. Ionov
  • A. I. Emel'yanov
  • T. K. Sedova
  • S. M. Balashov


Certain problems associated with the design and construction of and the results of work performed on large-scale safety test stands for water-moderated water-cooled, channel, and boiling-water vessel reactors are presented. It is shown that inadequate adherence to simulation principles in experimental setups can result in incorrect results concerning the effectiveness of emergency core-cooling systems in reactors.

The results obtained in this country and abroad show that a successful design, technological effectiveness, and reliability of all components of the model of a reactor core prevent the loss of time and resources in performing experimental programs. For example, such losses resulting from frequent failures of fuel element simulators, which are technologically highly effective and critical components of test stands, can be eliminated.


Critical Component Vessel Reactor Element Simulator Fuel Element Experimental Program 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.


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  1. 1.
    J. Adams, D. Batt, and V. Berta, “Influence of LOFT PWR transient simulations on thermal-hydraulic aspects of commercial PWR safety,” Nucl. Safety, 27, No. 2, 179–192 (1986).Google Scholar
  2. 2.
    S. Levy, “The important role of thermal hydraulics in 50 years of nuclear power applications,” Nucl. Eng. Design, 149, Nos. 1- 3, 1–10 (1994).Google Scholar
  3. 3.
    P. Weiss, R. Emmerling, R. Hertlein, and J. Leibert, “UPTF experiment refined PWR LOCA thermal-hydraulic scenarios: conclusions from a full-scale experimental program,” ibid., 149, Nos. 1- 3, 333–347 (1999).Google Scholar
  4. 4.
    H. Classer and H. Karwart, “Contribution of UPTF experiments to resolve some scale- up uncertainties in countercurrent two-phase flow,” ibid., 145, Nos. 1- 3, 63–84 (1993).Google Scholar
  5. 5.
    B. I. Nigmatullin, E. N. Videneev, and V. V. Zemlyanukhin, “Experimental setups for simulating accidents with a small leak in VVÉR-type reactors,” Teploénergetika, No. 12, 24–28 (1988).Google Scholar
  6. 6.
    B. G. Gordon, “Large-scale model of a nuclear power plant with a VVÉR reactor,” in: Operation and Repair of Equipment in a Nuclear Power Plant. Series on Power Engineering and Electrification. Express Information, Moscow (1986), No. 2, pp. 11–15.Google Scholar
  7. 7.
    B. I. Nigmatullin and S. M. Balashov, “Questions concerning the construction of a model for a reactor in a full-scale VVÉR safety stand,” in: International Conference “Thermal Physics-95,” Obninsk, November 21- 24, 1995.Google Scholar
  8. 8.
    S. M. Balashov, V. V. Zorichev, and A. S. Kon'kov, “Experience in using powerful electric heaters in thermophysical experiments,” in: International Conference “Thermal Physics-95,” Obninsk, May 26- 29, 1998, Vol. 1. pp. 372–380.Google Scholar
  9. 9.
    J. Blaisdell, PWR FLECHT Final Report, WCAP-7665 (1971).Google Scholar
  10. 10.
    M. Majed, G. Norback, P. Wiman, et al., Experience Using Individually Supplied Heating Rods in Critical Power Testing of Advanced BWR Fuel, NURETH-7 (1998), pp. 2608–2920.Google Scholar
  11. 11.
    S. M. Balashov, A. S. Kon'kov, and V. V. Zemlyanukhin, “Experimental investigation of repeated flooding in VVÉR fuel rod assemblies,” Teploénergetika, No. 6, 26–32 (1999).Google Scholar
  12. 12.
    V. V. Lozhkin, O. A. Sudnitsyn, and B. I. Kulikov, “Results of an experimental investigation on repeated cooling on fuel assembly models of a VVÉR reactor with flooding from below,” in: International Conference “Thermal Physics-95,” Obninsk, May 26- 29, 1998, Vol. 1. pp. 389–399.Google Scholar
  13. 13.
    Status Report to the Advisory Committee on Reactor Safeguards in the Matter of Westinghouse Electric Company ECCS Upper Head Injection Evaluation Model Conformance to 10 CFR50, Appendix K (1976).Google Scholar
  14. 14.
    Yu. A. Bezrukov, S. A. Logvinov, S. V. Levchuk, et al., “Construction of a first-scale model of a cassette of a VVÉR-440 reactor for investigating the core temperature at the repeated flooding stage,” in: International Conference “Thermophysics-82,” Karlovy-Vary, May 4- 8, 1982, Vol. 1, pp. 124–128.Google Scholar
  15. 15.
    V. V. Lozhkin, O. A. Sudnitsyn, and B. I. Kulikov, “Results of an experimental investigation of repeated cooling on VVÉR reactor fuel assembly models with flooding from above and combined flooding,” in: International Conference “Thermophysics-98,” Obninvk, May 26- 29, 1998, Vol. 1, pp. 381–388.Google Scholar
  16. 16.
    S. M. Balashov and B. G. Gordon, “Experimental investigation of repeated flooding of a VVÉR-440 cassette model,” in: Thermohydraulic Processes in Nuclear Power Plant Equipment, Énergoatomizdat, Moscow (1986), pp. 21–27.Google Scholar
  17. 17.
    H. Tuomisto, “Large-scale air/water flow tests for separate effects during LOCAs in PWR,” in: Proceedings of Specialist Meeting on Small Break LOCA Analyses on LWR, Pisa, Italy, June 23- 27, 1985, Vol. 1, pp. 149–162.Google Scholar
  18. 18.
    B. A. Gabaraev, V. N. Smolin, and S. L. Solov'ev, “Principles for building stands-models for verifying the thermohydraulic codes for calculating power reactors,” In: Thermophysical Aspects of Nuclear Power Plant Safety. International Conference “Thermophysics-2001,” Obninsk, May 29- 31, 2001, pp. 48–50.Google Scholar
  19. 19.
    V. N. Smolin, V. P. Shishov, A. I. Emel'yanov, et al., “Electric heaters for thermophysical models of fuel elements,” At. Énerg., 89, No. 6, 497–500 (2000).Google Scholar
  20. 20.
    V. N. Smolin and S. M. Balashov, “Experiments in developing and using fuel element simulators for thermophysical studies,” In: International Conference “Thermophysics-95,” Obninsk, November 21- 24, 1985.Google Scholar
  21. 21.
    Yu. I. Mityaev, Yu. I. Tokarev, I. N. Sokolov, et al., “New-generation nuclear power plants with enhanced-safety boiling water Bessel reactors,” At. Énerg., 73, No. 1, 13–19 (1992).Google Scholar

Copyright information

© Plenum Publishing Corporation 2002

Authors and Affiliations

  • V. N. Smolin
    • 1
  • V. P. Shishov
    • 1
  • V. I. Grachev
    • 1
  • A. I. Ionov
    • 1
  • A. I. Emel'yanov
    • 1
  • T. K. Sedova
    • 1
  • S. M. Balashov
    • 1
  1. 1.Federal State Unitary EnterpriseN. A. Dollezhal' Scientific-Research and Design Institute for Power EngineeringRussia

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