Fast Reactors and Accelerator Driven Systems Knowledge Base

Conference Article: The BOR-60 loop-channel design for testing the BREST reactor fuel

V.N. Efimov, I.Yu. Zhemkov, N. Kozolup, V.I. Polyakov, V.T. Stepanov, Yu.E. Stynda, V.V. Orlov, A.I. Filin, A.G. Sila-Novitski, A.A. Pikalov

Abstract

The paper is devoted to development and calculated substantiation of the design of the autonomous lead?cooled loop (AILCL) intended for testing the fuel pins prototypes in the BOR?60 reactor for the BREST?OD?300 reactor. The design features of the loop, its characteristics, instrumentation are considered. The auxiliary systems required to provide the loop operation are described. The main neutron?physical and thermohydraulic characteristics of the loop are presented. The basic operating conditions of the loop are shown. Great consideration is given to analysis of faults and failures affecting the loop serviceability and reactor safety. It is shown that the BOR?60 reactor safety is provided under all normal modes and postulated failures.

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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:77-96