THERMOHYDRAULIC STUDIES IN JUSTIFICATION THE STEAM GENERATORS OF REACTOR PLANTS WITH LIQUID METAL COOLING CARRIED OUT AT IPPE

EDN: FGSSFG

Authors & Affiliations

Grabezhnaya V.A., Mikheyev A.S., Alekhin A.V.
A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk, Russia

Grabezhnaya V.A. – Leading Researcher, Cand. Sci. (Tech.). Contacts: 1, pl. Bondarenko, Obninsk, Kaluga region, Russia, 249033. Tel.: +7 (484) 399-86-86; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Mikheyev A.S. – Senior Researcher.
Alekhin A.V. – Lead Systems Engineer, A.I. Leipunsky Institute for Physics and Power Engineering.

Abstract

Since the middle of the last century power units of nuclear power plants with fast reactors cooled by liquid metal have been created, built, and operated in the world. Steam generator (SG) is one of the main elements of such power units. For more than half a century in the SSC RF – IPPE at the SPRUT test facility on SG models work is carried out on the thermal hydraulic justification of steam generators of various reactor plants heated by different liquid-metals. Single-tube models of sodium-heated steam generators were tested to justify the evaporators of power units BN-350, BN-600; lead-heated multi-tube models were tested to justify the BREST-OD-300 SG, as well as single-tube models of sodium-heated steam generators were tested for new generation reactors of the BN-type. The main purpose of SG testing is to confirm the design characteristics, substantiate the thermohydraulic stability of steam generating channels when operating at specified mode parameters.
This paper gives a brief description of the tested SG models, presents the main results obtained during the tests. It is shown that during the tests, special attention was paid to the identification of modes with thermohydraulic instability at the operation of the steam generating channel, the occurrence of which significantly reduces the service life of the SG.

Keywords
steam generator, BN-350, BN-600, BREST-OD-300, sodium, lead, water, one-tube model, fragment model, thermohydraulic stability, heat transfer, longitudinal flow, transverse flow, basic oxygen mode, lead on the oxygen saturation line

Article Text (PDF, in Russian)

References

UDC 524.4:621.39.58

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2022, issue 2, 1:15