Atomic Energy

, Volume 77, Issue 2, pp 596–600 | Cite as

Possibility of using tube shell tables for calculating the heat-transfer crisis in rod bundles of water-cooled reactors

  • Yu. M. Zhukov


The results of the comparison of the procedure of [2] to the experimental data on the crisis of heat transfer in bundles rod fuel elements from the domestic and foreign databases show the following:

The formula of the Russian Scientific Center "Kurchatovskii institut" has much better statistical characteristics than Groeneveld's method in generalizing arrays of points on which the basis of the experimental generalization of the thermotechnical safety of domestic VVÉR reactors is based. The use of a shell table for a tube (AECL-86 variant) at low pressure (2–5 MPa) results in overestimation ofqcalc/qexp by up to a hundred of percent. At pressures above 12 MPa, however, the calculation is on the average 1.5 times lower than the experimental value.

For simulators modeling the conditions for the appearance of a heat-transfer crisis in PWR and BWR cores, the method of [2] describes well an array of 11,000 points in the Columbia University database. In the case of a triangular arrangement of rods, however, there is a large discrepancy (by up to 40%) between the calculation and experiment. The additional correction [3] of the constantK2 for bundles did not compensate the systematic underestimation by the calculation (by 37%) for a triangular arrangement of the simulators in the presence of axial nonuniformity of energy release.

Experience in operating domestic power reactors over a period of many years shows that the modern database and accurate empirical correlation make it possible to perform a reliable calculation of the admissible thermal power of rod bundles in a wide range of geometric and operating parameters, spacing methods, and energy-release fields without using the additional procedure of making the transition from a pipe to a bundle.


Power Reactor Fuel Element Spacing Method Experimental Generalization Additional Correction 
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Literature Cited

  1. 1.
    V. P. Bobkov, O. A. Zyatnina, N. V. Kozina, et al., "Crisis of boiling in channels with different geometry of the transverse cross section," Preprint FÉI-2314 (1993).Google Scholar
  2. 2.
    D. Groeneveld, S. Cheng, and T. Doan, "1986 AECL-UO critical heat flux lookup table," Heat Transfer Eng.,7, No. 1–2 (1986).Google Scholar
  3. 3.
    M. Lee and L. Liao, "An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the heat transfer research facility," Nucl. Tech.,105, 216–230 (1994).Google Scholar
  4. 4.
    V. N. Smolin and V. K. Polyakov, "Intensification of heat transfer in a rod bundle by means of local swirlers," in: TF-74 Proceedings "Investigations of critical heat fluxes in rod bundles," 1974, p. 307.Google Scholar
  5. 5.
    L. L. Kobzar' and A. I. Suslov, "Verification of the computational model for critical heat fluxes in the RELAP5/MOD3 program on the basis of databanks," Report No. IAÉ-32/1-1868-92.Google Scholar
  6. 6.
    M. Lee, K. Huang, and L. Liao, "Assessment of CHF correlation in thermal-hydraulic system analysis codes," Transaction ANS, No. 10–14, 645 (1991).Google Scholar
  7. 7.
    V. S. Osmachkin and N. N. Lystsova, "Comparison of experimental data on the conditions of the heat-transfer crisis in models of VVÉR-reactor fuel assemblies to the computational results obtained by the method developed by the Institute of Atomic Energy," Preprint No. IAÉ-2558, 1975. (1975).Google Scholar

Copyright information

© Plenum Publishing Corporation 1995

Authors and Affiliations

  • Yu. M. Zhukov

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