Nuclear Physics and Atomic Energy

▀ńň­ÝÓ ˘│šŔŕÓ ˛Ó ňÝň­Ńň˛ŔŕÓ
Nuclear Physics and Atomic Energy

  ISSN: 1818-331X (Print), 2074-0565 (Online)
  Publisher: Institute for Nuclear Research of the National Academy of Sciences of Ukraine
  Languages: Ukrainian, English, Russian
  Periodicity: 4 times per year

  Open access peer reviewed journal

 Home page   About 
Nucl. Phys. At. Energy 2018, volume 19, issue 2, pages 111-120.
Section: Atomic Energy.
Received: 16.03.2018; Accepted: 18.06.2018; Published online: 02.08.2018.
PDF Full text (ru)

Problems of calculation of heat transfer crisis in fuel assembles of water cooled reactors

G. I. Sharaevsky*, N. ╠. Fialko, L. B. Zimin, I. G. Sharaevsky

Institute for Safety Problems of NPP, National Academy of Sciences of Ukraine, Kyiv, Ukraine

*Corresponding author. E-mail address:

Abstract: Current problem of the ensuring reliability of the results of mathematical computer simulation of the operational modes of water-cooled nuclear reactors is considered in this article. An analysis of the adequacy of computer software systems, which are designed to calculate the main parameters of the safety of WWR reactors is performed The main focus is devoted to the methodology for determining the technological security of the active zones reactor plants settings, using the modern thermal-hydraulic codes. This calculation is based on determining the thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. The results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of computer codes are introduced. Particular attention is paid to the analysis of experimental and calculated data, by the definition of burnout in the fuel rods assembled elements. The basic directions of perfection of the modern thermal-hydraulic codes to improve the reliability of determination of thermophysical parameters of safety for the water-cooled nuclear reactors were considered.

Keywords: water-cooled reactors, parameters of safety, heat-hydraulic codes, heat transfer crisis.


1. └.V. Nosovsky et al. Thermophysics of NPP Resource (Chernobyl: Institute for Safety Problems of NPP NAS of Ukraine, 2017) 624 p. (Rus)

2. └.└. ╩liuchnikov et al. Thermophysics of NPP Safety (Chernobyl: Institute for Safety Problems of NPP NAS of Ukraine, 2010) 484 p. (Rus)

3. Yu.└. ╠igrov, S.L. Solovĺev. Thermohydraulic calculation codes of a new generation. General characteristics and prospects of development. Thermophysics (╬bninsk, 2001) p. 13. (Rus)

4. B.I. Nigmatulin, ╬.I. ╠elikhov, S.L. Solovĺev. State and development of domestic system thermal-hydraulic codes for modeling emergency and non-stationary processes at NPPs with WWER. Ďňploenergetika 3 (2001) 17. (Rus)

5. V.└. ┼fimov, D.P. Ďtrutnev, L.D. ╠Órchenko. Studying the crisis of boiling water in bundle rods. In: Heat Transfer, Hydrodynamics and Thermophysical Properties of Substances (╠oskva: Nauka, 1968) p. 60. (Rus)

6. V.I. Ďolubinsky, ┼.D. Domashev. The heat transfer crisis at boiling in bundle rods. Thermophysics and Heat Engineering 37 (1979) 3. (Rus)

7. V.I. Ďolubinsky, ┼.D. Domashev. On the causes of the discrepancy between the experimental data on the heat transfer crisis during boiling in channels. In: Heat and Mass Transfer in Liquids and Gases (Kyiv: Naukova dumka, 1984) p. 3. (Rus)

8. Yu.└. Bezrukov et al. Investigation of critical heat fluxes in beam beams for WWER-type reactors. Ďňploenergetika 2 (1976) 80 (Rus)

9. └.P. ╬rnatsky, L.F. Glushchenko, ┼.╠. ╠Óňvsky. Critical heat fluxes in steam-generating pipes in the region of small underheating and vapor content. Ďňploenergetika 8 (1971) 74. (Rus)

10. └.P. ╬rnatsky et al. Generalization of the results of the study of heat transfer crises in annular channels using input parameters. Thermophysics and Heat Engineering 26 (1974) 54. (Rus)

11. V.└. Kapustin et al. Experimental research at the stand of the I.V. Kurchatov KS IAE of critical thermal loads in full-scale models of fuel assemblies of the VVER-440 reactor. In: Investigation of Critical Heat Fluxes in Beams in Stationary and Non-Stationary Modes of Heat Exchange (╠oskva: I└E, 1974) p. 99. (Rus)

12. ┬.W. LeTourneau, S.I. Green. Critical heat flux and pressure drop tests with parallel upflow of high pressure in bundles of twenty 1/2-in. rods. Nucl. Sci. and Eng. 43(1) (1971) 90.

13. F. Luchini, V. Marinelli. Experimental data on burnout in a simulated BWR fuel bundle. Nucl. Eng. and Design 31(3) (1974) 371.

14. └.└. Kliuchnikov et al. Thermophysics of Active Zones Reliability (Chernobyl: Institute for Safety Problems of NPP NAS of Ukraine, 2015) 772 p. (Rus)

15. R.T. Lakhi, B.S. Shiralkar, D.V. Radklif. The distribution of mass velocity and enthalpy in a beam of rods for single-phase and two-phase flows. Trudy amerikanskogo obshchestva inzhenerov-mekhanikov. Ser. Đ 93(2) (1971) 64. (Rus)

16. Recommendations for calculating the heat transfer crisis in boiling water in round pipes. Preprint of the Institute of Computational Technologies NAS of USSR 1-57 (╠oskva, 1980) 67 p. (Rus)

17. V.N. Smolin, V.K. Poliakov. A method for calculating the heat transfer crisis at the boiling point of a coolant in rod assemblies. Proc. of the seminar TPh-78. Thermophysical studies to ensure the reliability and safety of nuclear reactors of water-water type. Vol. 2 (Budapesht, 1978) p. 475. (Rus)

18. V.N. Smolin, V.K. Poliakov. Critical heat flux in the case of longitudinal flow past a beam of rods. Ďňploenergetika 4 (1967) 54. (Rus)

19. Z.L. Miropolĺsky, A.G. Semin, ╠.N. Vinogradova. Statistical regularities in the study of the heat transfer crisis. Ďňploenergetika 7 (1969) 49. (Rus)