Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: Experimental and computational study on core thermohydraulics of BREST-type reactors (lead cooling)

A.D. Efanov, A.V. Zhukov, Ju.A. Kuzina, A.P. Sorokin, V.P. Smirnov, A.G. Sila-Novitski

Abstract

Experimental studies of heat-transfer coefficients and temperature fields of fuel rods for the BREST-type lead-cooled reactors have been performed, using thermo-hydraulic models which consist of fuel subassemblies with square rod arrangement and use an eutectic alloy sodium-kalium as a coolant. In the results obtained from the experimental models with three different pitches, thermohydraulic nonuniformity appears along perimeters of rods located in the energy release area due to different coolant heating around the rods, and it leads to the reduction of heat transfer coefficients. In addition, a computational code has been developed to evaluate thermohydraulic performances of the BREST-300 reactor core. Code verification using the experimental data showed a satisfactory agreement. All these experimental and computational studies will contribute to thermohydraulic substantiation of BREST-type reactor cores.

view the full text of this article (16 pages, format: PDF, size= 431kB)

key words: Fast Neutron Spectrum Systems, Nuclear Technology
Reference:
Advisory Group Meeting on “Design and Performance of Reactor and Sub-critical Blanket Systems with Lead and Lead-Bismuth as Coolant and/or Target Material” , Moscow, Russian Federation, 23-27 October 2000
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--1348, pp:106-121