TESTS OF MODELS OF BREST-OD-300 REACTOR FUEL ELEMENTS IN AN AUTONOMOUS LEAD-COOLED CHANNEL OF A BOR-60 REACTOR

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Abstract

A special feature of the BREST-OD-300 reactor that is now being designed is that it employs a container-type heat-conducting fuel element with mixed uranium–plutonium mononitride fuel, a lead sublayer, and an expansion volume at the top to collect gaseous products. The fuel elements are arranged in a square array with a wide spacing and are spaced by laminated spacing lattices.

The substantiation of the technical solutions adopted for the construction of the reactor fuel elements and fuel assemblies, specifically, the combined effect of the coolant and heat loads on the fuel-element cladding and the spacing lattices, led to the choice of the BOR-60 sodium-cooled fast research reactor as an experimental base and required the development and construction of an autonomous lead-cooled channel loaded into a cell through a passage in the rotatable plugs of the reactor. The channel was tested for two microruns with the BOR-60 reactor operating at 45 MW. The lead temperature at the fuel assembly entrance was 595°C, the working temperature of the cladding was ≤658°C, the damaging dose was 6.5 displacements/atom, and the fuel burnup was 0.44% h.a. Analysis of the activity of the gas and the lead showed that the fuel elements are sealed. Post-reactor studies have been conducted since August 2002.

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