Tests of fuel elements with uranium-plutonium nitride fuel in an IGR pulsed reactor

Kaplienko A.V. Lemekhov V.V. Cherepnin Y.S. Moiseyev A.V. Zhirnov A.P. Ivanyuta A.N. Rozhdestvenskiy I.M. Loginov D.Y. Mezhina Y.R. Izhutov A.L. Zvir Y.A. Shevlyakov G.V. Volkova I.N. Batyrbekov Y.G. Baklanov V.V. Korovikov A.G. Kotlyar A.N. Miller A.A. Irkimbekov R.A. Vurim A.D.
September 2023 Springer

Atomic Energy
2023 #134 Issue 5-6 275 - 282 pp.

Fuel elements with mixed uranium-plutonium nitride fuel were tested in an IGR reactor to justify their application in a BREST-OD-300 reactor. The mid-radial enthalpy limit for the fresh mixed nitride fuel as experimentally determined during IGR launching with fast reactivity input amounted to 167 cal/g. The maximum temperature of 1000 °C and its maintenance for 100 s, comprising one of the design limits for the fuel cladding temperature, was experimentally confirmed. The main results of the performed experiments and post-irradiation studies are analyzed.

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JSC NIKIET, Moscow, Russian Federation
JSC “RIAR”, Dimitrovgrad, Russian Federation
RSE NNC RK, Kurchatov, Kazakhstan

JSC NIKIET
JSC “RIAR”
RSE NNC RK

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