Abstract
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.
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REFERENCES
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).
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).
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).
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).
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).
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.
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.
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.
J. Blaisdell, PWR FLECHT Final Report, WCAP-7665 (1971).
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.
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).
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.
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).
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.
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.
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.
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.
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.
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).
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.
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).
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Smolin, V.N., Shishov, V.P., Grachev, V.I. et al. Some Problems of Large-Scale Safety Test Stands. Atomic Energy 92, 367–372 (2002). https://doi.org/10.1023/A:1019987225393
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DOI: https://doi.org/10.1023/A:1019987225393