Authors & Affiliations

Ivanov E.F., Privezentsev V.V., Sorokin A.P., Hafizov R.R., Trufanov A.A.
A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia

Privezentsev V.V. - Leading researcher of Division of Nuclear Power Plant Safety Cand. Sci. (Tech.), A.I. Leypunsky Institute for Physics and Power Engineering. Contacts: 1, sq. Bondarenko, Obninsk, Kaluga reg., Russia, 249033, Tel.: (484) 399-56-57, e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it.
Sorokin A.P. – Deputy Director of Division of Nuclear Power Plant Safety, Dr. Sci (Tech.), A.I. Leypunsky Institute for Physics and Power Engineering.
Ivanov E.F. – Leading Researcher of Division of Nuclear Power Plant Safety, Cand. Sci. (Tech.), A.I. Leypunsky Institute for Physics and Power Engineering.
Hafizov R.R. – Scientific Researcher of Safety Department,
Trufanov A.A. – Deputy Director General, Director of Safety Department, A.I. Leypunsky Institute for Physics and Power Engineering.

Abstract

Calculation modeling of ULOF type accidents for sodium fast reactor indicates coolant boiling initiation in the reactor core. Significant impact on calculation results is provided by two-phase flow calculation model used in the code and which requires experimental verification. In order to eliminate the possibility of development accidental situations leading to fuel pins destruction, it was proposed implementation of sodium cavity above reactor core. On the facility created at SSC RF – IPPE were obtained experimental data on boiling heat exchange in the fast reactor fuel subassembly model with sodium cavity under natural and forced coolant circulation conditions. The possibility of long-term fuel subassembly cooling at heat fluxes from 140 to 170 kW/m2 was shown under natural and forced convection respectively. The data obtained used torefine the sodium boiling calculation model and verification of the COREMELT code.

Keywords
fast reactor, experimental investigation, subassembly model, fuel element,heat flux, natural and forced circulation, sodium boiling, two-phase flow, flow regime, heat transfer

Article Text (PDF, in Russian)

References

UDC 621.039.526.034+621.039.546.8:536.26

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2016, issue 5, 5:9