Abstract
Experience with the operation of nuclear power plant (NPP) steam generators (SGs) has shown that the main factor determining the technical condition and actual service life of a steam generator (SG) is the condition of the heat-transfer tubes, which belong to the parts of a steam generator that cannot be replaced or restored. Corrosion defects in heat-transfer tubes initiate and grow during operation of power units. This can lead to depressurization of the primary circuit of the nuclear power plant and primary-to-secondary leakage during on-load operation of the reactor unit. When this leakage in the steam generator exceeds a certain standard value, the power unit is shut down for unscheduled inspection of heat-transfer tubes for integrity. To ensure safe operation of the power unit and to avoid unscheduled shutdown of the reactor unit, the heat-transfer tubes are subjected to nondestructive examination, the results of which are used to plug faulty tubes. This paper presents the results of the development and validation of the thermohydraulic CFD model of the NPP primary circuit in the PGV-1000M steam generator, taking into account the three-dimensional effects in the steam generator headers. To simulate the thermohydraulic processes in heat-transfer tubes of a steam generator heat-transfer tubes in the STAR-CCM + code, a one-dimensional flow and heat-transfer model describing changes in the thermohydraulic parameters of the coolant only along the length of the tubes is implemented in the STAR-CCM+ code. Thermohydraulic calculations of a steam generator with several plugged heat-transfer tubes and the analysis of the effect of their location on the thermohydraulic characteristics of the steam generator under rated operating conditions of the reactor unit have been performed. The results of optional calculations were obtained using the CFD model of the primary SG circuit when varying the number of plugged heat-transfer tubes and their location. The effect of the number of plugged tubes and their location on hydrodynamics and heat transfer in the considered SG has been analyzed.
Similar content being viewed by others
Notes
In this study, deposits on the SG heat-transfer tube outside were neglected.
REFERENCES
N. B. Trunov, V. S. Popadchuk, S. E. Davidenko, and R. Yu. Zhukov, “Pressing problems of managing the service life of tube bundles used in steam generators at nuclear power stations equipped with VVER reactors,” Therm. Eng. 57, 361–369 (2010).
V. Yu. Volkov, L. A. Golibrodo, A. A. Krutikov, O. V. Kudryavtsev, Yu. N. Nadinskii, A. T. Nechaev, and A. P. Skibin, “Multi-scale problems of heat and mass transfer in nuclear industry,” Vestn. YuUrGU, Ser. Vychisl. Mat. Inf. 6, 60–73 (2017). https://doi.org/10.14529/cmse170405
N. B. Trunov, S. A. Logvinov, and Yu. G. Dragunov, Hydrodynamic and Thermal Chemical Processes in Steam Generators of VVER-Based Nuclear Power Plants (Energoatomizdat, Moscow, 2001) [in Russian].
B. I. Lukasevich, N. B. Trunov, Yu. G. Dragunov, and S. E. Davidenko, Steam Generators of VVER Reactor Plants for Nuclear Power Plants (Akademkniga, Moscow, 2004) [in Russian].
M. V. Teslenko, “Three-dimensional representation of heat-exchanging tube flaws in steam generators,” Vopr. At. Nauki Tekh., Ser. Obespechenie Bezopasnosti AES, No. 31, 62–71 (2012).
RD EO 1.1.2.16.0157-2009. Norms of Damage (Blanking-Off Criteria) for Heat-Transfer Tubes of Steam Generators Used as Part of Reactor Plants at VVER-Based Nuclear Power Plants (Energoatom, Moscow, 2009).
V. F. Strizhov, M. A. Bykov, A. Ye. Kiselev A. V. Shishov, A. A. Krutikov, D. A. Posysaev, and D. A. Mustafina, “Development of a 3D model of tube bundle of VVER reactor steam generator,” in Proc. Workshop on Experiments and CFD Codes Application to Nuclear Reactor Safety (XCFD4NRS), Grenoble, France, Sept. 10–12, 2008 (Nuclear Energy Agency, Organisation for Economic Co-operation and Development, Paris, 2008).
K. S. Dolganov and A. V. Shishov, “Cross-verification of one- and three-dimensional models for VVER steam generator,” in Proc. Workshop on Experiments and CFD Codes Application to Nuclear Reactor Safety (XCFD4NRS), Grenoble, France, Sept. 10–12, 2008 (Nuclear Energy Agency, Organisation for Economic Co-operation and Development, Paris, 2008).
D. A. Posysaev, O. V. Kudryavtsev, A. P. Skibin, A. V. Shishov, M. A. Bykov, and N. B. Trunov, “Calculation of the hydrodynamics of the structure of the primary circuit of the PGV-1500 steam generator using CFD codes,” in Proc. 7th Int. Seminar on Horizontal Steam Generators, Podolsk, Oct. 3–5, 2006 (Gidropress, Podol’sk, 2006).
USER GUIDE STAR-CCM+, Version 13.04 (SIEMENS PLM Software, 2018).
I. E. Idel’chik, Handbook of Hydraulic Resistances (Ripol Klassik, Moscow, 2013) [in Russian].
R. Fernandez-Prini and R. B. Dooley, Release on the IAPWS Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam (International Association for the Properties of Water and Steam, Erlangen, Germany, 1997).
Author information
Authors and Affiliations
Corresponding author
Additional information
Translated by T. Krasnoshchekova
Rights and permissions
About this article
Cite this article
Volkov, V.Y., Goliborodo, L.A., Krutikov, A.A. et al. Modeling of Thermohydraulic Processes in a Steam Generator with Several Plugged Heat-Transfer Tubes. Therm. Eng. 69, 77–86 (2022). https://doi.org/10.1134/S0040601522020070
Received:
Revised:
Accepted:
Published:
Issue Date:
DOI: https://doi.org/10.1134/S0040601522020070