Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: Experimental studies of BREST-OD-300 reactor characteristics on BFS facilities

I.P. Matveenko, A.M. Tsyboulia, G.N. Manturov, M.Y. Semyonov, V.N. Kocsheyev, V.S. Smirnov, A.V. Lopatkin, V.G. Muratov, P.N. Alexeyev

Abstract

The BFS-77-1 critical experiments were performed to validate and verify the software and cross section data base aiming at confirming the neutronics analysis of lead cooled BREST-OD-300 fast reactor core. For the validation and verification, other experimental works have been also completed to justify software and cross section data base, and real accuracy of calculation of characteristics of the BREST-OD-300 fast reactor core. It was drawn from the analysis of current status of evaluated neutron data for lead that additional experiments are required in order to refine lead cross section values. Calculated characteristics of control rods and reactivity balance still require justification.

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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:50-63