MODELING OF HEAT EXCHANGE PROCESSES IN THE STEAM GENERATOR SGV-1000
Journal Title: Вчені записки Таврійського національного університету імені В. І. Вернадського. Серія: Технічні науки - Year 2018, Vol 29, Issue 3
Abstract
During the operation of steam generators working in conjunction with VVER-1000 nuclear power reactors of nuclear power plants, the problem of pipe cracking and the flow of primary circuit water inevitably arise, which causes scheduled preventive maintenance (PM). During PM, the flowing pipes are drowned, as a result of which, on the one hand, the heat exchange area decreases, and on the other hand, the velocity of the coolant increases. Thus, the problem of studying the dependence of the change in the temperature of the coolant of the primary circuit of the nuclear reactor, when the number of tubes in the steam generator of the nuclear power plant changes, becomes urgent. The effect of changing the water temperature of the primary circuit on the reactor power is estimated.
Authors and Affiliations
В. В. Ткаченко, К. В. Беглов, Е. О. Улицкая
Keywords
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