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Dimitrovgrad, Russia

Steinbruck M.,Karlsruhe Institute of Technology | Birchley J.,Paul Scherrer Institute | Boldyrev A.V.,RAS Nuclear Safety Institute | Goryachev A.V.,RIAR | And 10 more authors.
Progress in Nuclear Energy | Year: 2010

This paper gives an overview on the status of knowledge of high-temperature oxidation of the two zirconium alloys Zircaloy-4 and E110 with special emphasis on results obtained during the SARNET period. The tin-bearing alloy Zircaloy-4 and the niobium-bearing alloy E110 are the materials for cladding and structures used in pressurised water reactors of the Western-type and VVERs and RBMKs, respectively. Results of separate-effects tests, single-rod tests, and large-scale bundle experiments are summarised. Focus is directed to oxidation kinetics at high temperature, hydrogen release and absorption by the remaining metal, and behaviour during quenching. Both materials behave very similarly as long as the superficial oxide scales formed during oxidation are dense and protective. Main differences are seen in connection with breakaway oxidation which leads to enhanced oxidation and hydrogen uptake and thus embrittlement and possibly earlier failure of the cladding. The temperature range at which pronounced breakaway is observed is different for the two alloys. The status of modelling of oxidation kinetics, thermo-mechanical behaviour during cooldown and the influence of irradiation are discussed at the end of the paper. © 2009 Elsevier Ltd. All rights reserved. Source


Izhutov A.L.,RIAR | Iakovlev V.V.,RIAR | Novoselov A.E.,RIAR | Starkov V.A.,RIAR | And 7 more authors.
Nuclear Engineering and Technology | Year: 2013

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ~ 60%235U; the mini-rods were irradiated to an average burnup of ~ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%. Source

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