Conclusions
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 ofq calc/q exp 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 constantK 2 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.
This is a preview of subscription content, log in via an institution to check access.
Similar content being viewed by others
Literature Cited
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).
D. Groeneveld, S. Cheng, and T. Doan, "1986 AECL-UO critical heat flux lookup table," Heat Transfer Eng.,7, No. 1–2 (1986).
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).
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.
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.
M. Lee, K. Huang, and L. Liao, "Assessment of CHF correlation in thermal-hydraulic system analysis codes," Transaction ANS, No. 10–14, 645 (1991).
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).
Additional information
Institute of Nuclear Reactors, Russian Scientific Center "Kurchatovskii Institut." Translated from Atomnaya Énergiya, Vol. 77, No. 2, pp. 108–111, August, 1994.
Rights and permissions
About this article
Cite this article
Zhukov, Y.M. Possibility of using tube shell tables for calculating the heat-transfer crisis in rod bundles of water-cooled reactors. At Energy 77, 596–600 (1994). https://doi.org/10.1007/BF02407432
Received:
Issue Date:
DOI: https://doi.org/10.1007/BF02407432