Program and Results of Reactor Tests of Mixed Nitride Fuel for Fast Reactors

Abstract

In the Proryv project, mixed uranium-plutonium nitride fuel is to be used in the future nuclear power with fast reactors with sodium coolant (BN-1200) and with lead coolant (BREST). At present, experimental fuel elements and fuel assemblies with mixed nitride fuel are fabricated by means of a technology developed at the Bochvar All-Russia Research Institute for Inorganic Materials for testing in the BN-600 reactor and the MIR and BOR-60 research reactors. The program of reactor tests of the experimental fuel assemblies and some preliminary results of these tests are presented. World experience in research on mixed uranium-plutonium nitride fuel is limited to ~200 fuel elements, irradiated to maximum burnup ~18% h.a. at maximum linear power density 410–1300 W/cm [1, 2]. In our country, we have the results of the irradiation of ~1300 fuel elements with uranium nitride fuel in the BR-10 reactor [3]. Even though many properties of uranium nitride and mixed nitride fuel are close, the results of the research cannot fully serve as validation for the BREST and BN-600 fuel elements, for which the operating parameters determining their serviceability differ considerably from those of fuel elements with uranium nitride fuel. The results of the irradiation of four fuel elements with mixed nitride fuel to maximum burnup 12.1% h.a. in the BOR-60 reactor as part of a Russian–French experiment must also be taken into account [4]. Low swelling and gas-release and the corrosion damage to the cladding in this experiment were obtained for high-purity mixed nitride fuel fabricated from the initial metals with oxygen content <0.15 wt.% and carbon content <0.1 wt.% with pellet density 85%. The mixed nitride used in the experiment of [4] was characterized by elevated plutonium content 45 and 60 %, which could have promoted high creep rates and, correspondingly, lower stresses in the fuel-element cladding. For this reason, it is necessary to have information on its in-reactor behavior with the lower plutonium content provided in the fuel elements of the BREST and BN-1200 reactors. Analysis showed that it is necessary to obtain additional data on the pre-reactor and reactor properties of mixed uranium-plutonium nitride fuel. This requires reactor tests of BN-1200 and BREST type fuel elements in the BOR-60 and other research reactors in order to achieve statistically representative validation of the serviceability of mixed nitride fuel as part of full-scale experimental BN-600 fuel assemblies. A complex program of computational and experimental validation of using mixed uranium-plutonium nitride fuel in the BN-1200 and BREST-OD-300 reactors was developed as part of the Proryv project [5]. This program provides for pre-reactor investigations of the properties of fuel fabricated by the carbothermal synthesis technology and reactor tests of fuel elements in the BN-600 reactor and the MIR and BOR-60 research reactors [6]. The serviceability of fuel elements with