Authors & Affiliations
Grabezhnaya V.A., Mikheyev A.S., Alekhin A.V., Kryukov A.E., Tikhomirov A.A.
A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia
Mikheev A.S. – Senior Researcher.
Kryukov A.E. – Leading Thermal Physics Engineer.
Alekhin A.V. – Leading Systems Engineer.
Tikhomirov A.A. – Research Engineer 1 category, A.I. Leypunsky Institute for Physics and Power Engineering.
The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation – IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-tube modules with a longitudinal lead flow around a three-tube bundle of coiled tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters. At the second stage, tests of a standard steam generator model were carried out with lead flowing around 18 heat exchange tubes. In the multitube model, the downward movement of the heating coolant took place with the flow around the bundle of heat transfer tubes close to the transverse flow. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of changes in operating parameters, which are necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. During the tests in both models of the steam generator, there were no noises inherent in unstable operating modes of the circuit. No pulsations of water and steam temperature were found, respectively, in the inlet and outlet collectors. At high lead temperatures, the temperature of the superheated steam was always close to the lead inlet temperature. A series of works devoted to the study of heat transfer from the side of a lead coolant with a transverse flow around a package of heat exchange tubes in normal heat transfer modes and with freezing of lead has been completed. Studies have been carried out on the effect of oxygen concentration in lead on heat transfer.
steam generator, lead, water, coiled duct, bundle of heat exchange tubes, model, thermohydraulic stability, longitudinal flow, cross flow, heat transfer, basic oxygen mode, lead on the oxygen saturation line, lead freezing
Article Text (PDF, in Russian)
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Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2021, issue 2, 2:15