Series: Nuclear and Reactor Constants

since 1971

Русский (РФ)

ISSN 2414-1038 (online)

DOI: 10.55176/2414-1038-2019-1-132-151

Authors & Affiliations

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

Grabezhnaya V.A. – Leading Researcher, Cand. Sci. (Tech.), A.I. Leypunsky Institute for Physics and Power Engineering. Contacts: 1, pl. Bondarenko, Obninsk, Kaluga region, Russia, 249033. Tel.: +7(484) 399-42-97, e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Mikheyev A.S. – Senior Researcher, A.I. Leypunsky Institute for Physics and Power Engineering.


The study of heat transfer in spiral coiled tubes is of great interest in view of the widespread use of such channels in engineering practice, in particular, in nuclear power engineering in the form of steam generators at research reactors and nuclear power plants. In the projected BREST-OD-300 reactor facility (RF), a configuration of helical coiled tubes is considered as a steam generator.
Thermal hydraulic tests of the model steam generator RF BREST-OD-300 (version 2000) with helical coiled tubes with longitudinal coolant flow were carried out in SSC RF — IPPE at the SPRUT facility in 2011—2013 years. The test program was aimed to study heat transfer and thermal hydraulic stability of the steam generating tubes. Throughout the range of variation the regime parameters, regimes with a reversal circulation in the water loop have not been detected.
Despite the fact that the results of conducted tests on the steam generator model gave extensive information on heat transfer in different zones of the steam generating tube, however, the insufficient number of heat transfer tubes in the module (only three) does not allow to conclude that the full hydrodynamic stability of BREST RF steam generator in all possible modes of operational parameters. On the other hand, in a real construction the motion of the heating coolant is omitted with flow around the bundle of tubes close to the transverse flow. Therefore, insufficient reasoning for the transferring results obtained on a three-tube model to a full-scale steam generator served as the basis for testing of a multi-tube full-height fragment model of a reduced diameter one row of the tube bundle actual steam generator.
During the tests, there were no noises typical of the unstable operation of water loop. There were no oscillations of water and steam temperature, respectively, at the inlet to the collectors and out of the collectors. At high lead temperatures, the temperature of the superheated steam was always close to the inlet temperature of the lead.
The tests showed the absence of the thermohydraulic instability, as in the case of longitudinal and transverse coolant flows in the investigated modes of lead and water parameters. Other parameters being equal, the steam temperature at the outlet of the steam generating tube in case of transverse flow was higher than in the case of longitudinal flow.
The experimental data obtained during the testing are primarily necessary for the verification of codes that allow the correct calculation of the various operating modes of the BREST-OD-300 steam generator.

reactor, steam generator, heavy liquid coolant, lead, water, steam, helical coiled channel, bundle of heat transfer tubes, model of steam generator, experiment, thermohydraulic stability, longitudinal flow, transverse flow, the temperature profile

Article Text (PDF)


UDC 536.24.08

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2019, issue 1, 1:11